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Elia Merzari

Publications and source records attributed to Elia Merzari.

At least 19 recordsLinked to original sources

Super-Resolution Initialization of High-Fidelity CFD Simulations for Pebble-Bed Reactors

High-order CFD simulations provide detailed resolution of the heterogeneous interstitial flow in pebble-bed reactors, but their computational cost is high, especially during the initial flow-development period required to reach statistically stationary conditions. This work investigates the use of a Super-Resolution Graph Neural Network (SR-GNN) to improve the initialization of high-order NekRS simulations. Lower-order P = 2 velocity fields are used as inputs to reconstruct higher-order representations, which are then used as initial conditions for P = 7 restart simulations. The approach is evaluated using a 146-pebble bed at Re = 1000, Re = 2500, and Re = 5000, with pressure-drop convergence used as the main figure of merit. The SR-GNN models were trained using paired low- and high-order snapshots and were first evaluated through qualitative inference comparisons. High-order restart simulations showed that, for Re = 1000 and Re = 2500, the SR-GNN initialized cases produced pressure-drop histories similar to direct restarts from true P = 2 fields. For Re = 5000, however, the super-resolved field restart approached the statistically stationary P = 7 pressure-drop range faster than both the direct P = 2 restart and the reference P = 7 simulation initialized from a uniform velocity field. The trained Re = 5000 model was also applied to a larger 1568-pebble bed, demonstrating qualitative applicability of the workflow to a significantly larger packed-bed geometry. These results indicate that SR-GNN-based initialization is a promising strategy for reducing high-order flow-development cost, while also motivating further work on broader Reynolds-number and geometry generalization.

physics.flu-dyn

Multiphysics Modelling of the Molten Salt Fast Reactor using NekRS and the Fission Matrix Method

The Molten Salt Fast Reactor (MSFR) has the particularity that the coolant is also the fuel, which tightens the coupling between neutronics and thermal hydraulics as the fuel circulates through the primary system. Therefore, developing computational models to analyze the MSFR requires a multiphysics approach. In this paper, we propose developing a neutronic thermal-hydraulic computational model of the MSFR that uses a reduced-order model to solve the neutronics equations. The principal computational tool chosen for this purpose is the high-fidelity code Cardinal, a wrapping within the MOOSE framework that integrates the Computational Fluid Dynamics code NekRS and the Monte Carlo particle transport code OpenMC. However, we use the Fission Matrix (FM) Method to solve the neutronics equations instead of OpenMC. The FM method can perform fast and still accurate neutronics simulations. It relies on precalculated databases obtained through a Monte Carlo simulation.

physics.comp-ph

Analysis of Fission Matrix Databases using Temperature Profiles obtained from High-Fidelity Multiphysics Simulations

The Fission Matrix method is used to perform fast and still accurate neutronics simulations. It relies on precalculated databases obtained through a Monte Carlo simulation. To represent every state of the reactor, multiple databases are required. The actual state of the reactor is obtained from those databases. In this paper, we analyze the effect of the temperature profiles selected to construct the databases. To do so, the Molten Salt Fast reactor is selected. Two sets of databases are studied: the first uses temperature profiles obtained from high-fidelity Multiphysics simulations with Cardinal, and the second uses uniform temperature profiles. Results showed improved multiplication factor and fission source distribution when the temperature profiles used to generate the databases were similar to those expected when solving the fission matrix.

physics.comp-ph

Toward Developing Machine-Learning-Aided Tools for the Thermomechanical Monitoring of Nuclear Reactor Components

Proactive maintenance strategies, such as Predictive Maintenance (PdM), play an important role in the operation of Nuclear Power Plants (NPPs), particularly due to their capacity to reduce offline time by preventing unexpected shutdowns caused by component failures. In this work, we explore the use of a Convolutional Neural Network (CNN) architecture combined with a computational thermomechanical model to calculate the temperature, stress, and strain of a Pressurized Water Reactor (PWR) fuel rod during operation. This estimation relies on a limited number of temperature measurements from the cladding's outer surface. This methodology can potentially aid in developing PdM tools for nuclear reactors by enabling real-time monitoring of such systems. The training, validation, and testing datasets were generated through coupled simulations involving BISON, a finite element-based nuclear fuel performance code, and the MOOSE Thermal-Hydraulics Module (MOOSE-THM). We conducted eleven simulations, varying the peak linear heat generation rates. Of these, eight were used for training, two for validation, and one for testing. The CNN was trained for over 1,000 epochs without signs of overfitting, achieving highly accurate temperature distribution predictions. These were then used in a thermomechanical model to determine the stress and strain distribution within the fuel rod.

cs.LG

High-Fidelity Modelling of the Molten Salt Fast Reactor

The Molten Salt Fast Reactor (MSFR) is one of the six GEN-IV reactor designs. In the MSFR, the liquid fuel is the coolant, which moves throughout the primary circuit. This complex phenomenology requires multiphysics modeling. In the present paper, a model of the MSFR is developed in the multiphysics code Cardinal, considering neutronic-thermal hydraulic feedback and the transport of delayed neutron precursors (DNPs) and decay heat precursors (DHPs). OpenMC is used to solve neutronic equations, and NekRS is used to solve mass, momentum, energy, DNPs, and DHPs distribution. A RANS k-t turbulence model is used in NekRS. DNPs and DHPs are modeled using a convective-diffusion equation with modified source terms considering radioactive decay. Cardinal results showed a reasonable behavior for temperature, heat source, velocity, DNPs, and DHPs. However, the current limitations in OpenMC do not allow the modification of delayed neutron source locations. Ongoing efforts look to include this feature in future work to introduce DNP feedback in OpenMC.

physics.comp-ph

Multiphysics Modeling of SNAP 10A/2 Space Reactor with Cardinal

The SNAP 10/A nuclear-powered satellite was launched into space in 1965. The present work discusses the development of a coupled neutronic-thermal hydraulics model with the high-fidelity multiphysics code Cardinal for SNAP 10/A. A comparison of neutronic benchmarks between OpenMC and MCNP5 shows a difference of 304+/-70 pcm. The primary source of difference is that MCNP5 uses a combination of cross-section libraries ENDF/B-VI and ENDF/B-V for minor isotopes. At the same time, the present work utilizes ENDF/B-VIII.0. A comparison against experimental results showed a difference of 355 pcm. A coupled model of the SNAP 10/A reactor is then developed in Cardinal. Temperature results show reasonable agreement with reference calculations.

physics.comp-ph

Exascale Simulations of Fusion and Fission Systems

We discuss pioneering heat and fluid flow simulations of fusion and fission energy systems with NekRS on exascale computing facilities, including Frontier and Aurora. The Argonne-based code, NekRS, is a highly-performant open-source code for the simulation of incompressible and low-Mach fluid flow, heat transfer, and combustion with a particular focus on turbulent flows in complex domains. It is based on rapidly convergent high-order spectral element discretizations that feature minimal numerical dissipation and dispersion. State-of-the-art multilevel preconditioners, efficient high-order time-splitting methods, and runtime-adaptive communication strategies are built on a fast OCCA-based kernel library, libParanumal, to provide scalability and portability across the spectrum of current and future high-performance computing platforms. On Frontier, Nek5000/RS has achieved an unprecedented milestone in breaching over 1 trillion degrees of freedom with the spectral element methods for the simulation of the CHIMERA fusion technology testing platform. We also demonstrate for the first time the use of high-order overset grids at scale.

cs.CE

Study of Turbulence and Pressure Recovery in the Heat Pipe Vapor Flow Using the Spectral-Element Method

Heat pipes can efficiently and passively remove heat in nuclear microreactors. Nevertheless, the flow dynamics within heat pipes present a significant challenge in designing and optimizing them for nuclear energy applications. This work aims to explore the vapor core of heat pipes through comprehensive two- and three-dimensional simulations, with a primary focus on modeling the pressure recovery observed in the condenser section. The goal is to establish improved correlations for one-dimensional heat pipe codes. The simulations are validated against experimental data from a vapor pipe documented in the literature. The turbulence model is employed in the two-dimensional simulations through the open-source spectral-element code Nek5000. This model provides insights into pressure recovery within heat pipes with low computational cost. In addition, Large Eddy Simulations (LES) are used to capture turbulent flow features in a three-dimensional vapor pipe model, utilizing the code NekRS. Using LES is crucial for comprehending the impact of laminar-to-turbulent transition on pressure recovery. A simulation framework is created to model the heat pipe's vapor core, laying the foundation for an enhanced understanding of heat pipe behavior. The ultimate goal is to improve and optimize heat pipe designs, provide data to validate lower-fidelity approaches and enhance their performance in nuclear reactors.

physics.flu-dyn

Validation and Comparison of HI-STORM Overpack Thermal-Hydraulic Model with MOOSE and NekRS

Nuclear power is a significant source of electricity in the United States, but the average U.S. nuclear power plant is around 40 years old. Safe management of spent nuclear fuel (SNF) is a crucial aspect of the back end of the nuclear fuel cycle. SNF dry storage systems are increasingly popular as they represent an effective solution in this area, given the absence of a final disposal system. In particular, the spent fuel cask system (dry cask method) provides a feasible solution for maintaining SNF ($\sim$60 years) prior to the final disposal. The HI-STORM overpack and MPC-32 canister are the primary components of the HI-STORM 100 dry cask storage system. They remove heat from the system via natural circulation with no human intervention required. This characteristic provides passive heat removal while requiring little maintenance in dry cask storage systems. This project aims to validate and compare the capabilities of a thermal model of HI-STORM overpack developed using the Multiphysics Object-Oriented Simulation Environment (MOOSE) based on the author's previous study. MOOSE is an open-source framework developed by Idaho National Laboratory for multiscale, multiphysics simulations. This study will improve the capabilities of the thermal-hydraulic model of the HI-STORM dry cask storage system by producing high-fidelity results for the air circulation in the overpack. Large Eddy Simulations (LES) are performed using the open-source spectral element code NekRS, developed by Argonne National Laboratory (ANL), for simulating transitional and turbulent flows in complex geometries. NekRS will produce high-fidelity results for the HI-STORM overpack to assess the validity of the current thermal-hydraulic model.

physics.flu-dyn

Study on Laminar-Turbulent Transition in Square Arrayed Rod Bundles

The study of coolant flow behavior in rod bundles is of relevance to the design of nuclear reactors. Although laminar and turbulent flows have been researched extensively, there are still gaps in understanding the process of laminar-turbulent transition. Such a process may involve the formation of a gap vortex street as the consequence of a related linear instability. In the present work, a parametric study was performed to analyze the spatially developing turbulence in a simplified geometry setting. The geometry includes two square arrayed rod bundle subchannels with periodic boundary conditions in the cross-section. The pitch-to-diameter ratios range from 1.05 to 1.20, and the length of the domain was selected to be 100 diameters. No-slip condition at the wall, and inlet-outlet configuration were employed. Then, to investigate the stability of the flow, the Reynolds number was varied from 250 to 3000. The simulations were carried out using the spectral-element code Nek5000, with a Direct Numerical Simulation (DNS) approach. Data were analyzed to examine this Spatio-temporal developing instability. In particular, we evaluate the location of onset and spatial growth of the instability.

physics.flu-dyn

Toward Direct Numerical Simulation of Turbulent and Transitional Flow in Hexagonal Subchannels for Helium Conditions

Understanding the coolant thermal hydraulics in rod bundles is essential to the design of nuclear reactors. However, flows with low Reynolds numbers present serious modeling challenges, especially in heat transfer and natural convection. They are difficult to analyze through standard Computational Fluid Dynamics (CFD) tools. High-fidelity simulations, such as Direct Numerical Simulations (DNS), can provide invaluable insight into flow physics, supporting experiments in developing a deeper understanding and eventually enabling the accurate simulation of this class of flows. Data generated from these high-fidelity methods can then be used to benchmark available turbulence models and deliver cheap, faster running methods. In the present work, the convective heat transfer in hexagonal subchannels was studied through a DNS approach, using the high-order spectral element method code Nek5000, developed at Argonne National Laboratory. First, the geometric model composed of two hexagonal arrayed rod bundle subchannels with a pitch-to-diameter ratio of 1.5 is built, and then, the mesh is generated. These unusually high P/D and low Reynolds numbers represent conditions of interest for gas-fast reactors (GFRs). To our knowledge, there is no available dataset in these conditions. In this work, we detailed the development of the numerical benchmark and a series of preliminary LES simulations. Periodic boundary conditions are applied in the streamwise and spanwise directions and non-slip boundary conditions at the wall. Four cases are studied, with Reynolds of Re=2500, 5000, 7500, and 10000. All calculations have been performed with the Prandtl number of 0.61, corresponding to helium conditions of interest for Gas Fast Reactor applications. The results are analyzed with a polynomial-order convergence study, and the Reynolds stresses, and the turbulent kinetic budgets are presented and discussed.

physics.flu-dyn

High-fidelity simulation of pebble beds: Toward an improved understanding of the wall channeling effect

Wall channeling is a phenomena of interest for Pebble Bed Reactors (PBRs) where flow is diverted into high-porosity regions near the wall. This diversion of flow can have a significant impact on maximum fuel temperatures and core bypass flow. Porous media models that are currently used to model PBRs for design scoping and transient simulation are lacking in their capabilities to model the wall channel effect. Recent efforts at Penn State have produced an improved porous media pressure drop equation that is more capable of modeling the velocity variations caused by the wall channel effect in a porous media model. Several pebble beds were divided into concentric rings of $0.05D_{peb}$, and average flow quantities and porosities were extracted for the ring. A correlation between the form loss coefficient and the local ring porosity was found, allowing for the addition of a correction factor to the form loss term of the KTA equation. The developed correlation was purely empirical, and thus a more thorough understanding of the underlying flow phenomena is desired. This study investigates geometric and flow features that can explain the observed correlation between the form coefficient and the local porosity that was used to generate the improved pressure drop equation. The solid surface area to volume ratio $S_v$ along with the production of Turbulent Kinetic Energy (TKE) is analyzed. A relationship between $S_v$ and the local porosity and an inverse relationship between the negative TKE production and the local porosity were found, pointing to the idea that inertial effects caused by different pore geometry in each ring contribute to the variation of the form constant with the local porosity.

physics.flu-dyn

On the Impact of Aspect Ratio and Other Geometric Effects on the Stability of Rectangular Thermosiphons

Single-phase natural circulation thermosiphon loops have been attracting increased interest as they represent the prototype of passive safety systems. However, the stability properties of thermosiphon loops, which can affect and compromise their functionality, are still actively investigated. Traditionally, the stability analysis of thermosiphon loops has been simplified to one-dimensional (1D) calculations, on the argument that the flow would be mono-dimensional when the diameter of the pipe D is orders of magnitude smaller than the length of the loop Lt. However, at lower Lt/D ratios, rectangular thermosiphon loops show that the flow presents 3D effect, which also has been confirmed by stability analyses in toroidal loops. In this paper, we performed a series of high-fidelity simulations using the spectral element code Nek5000 to investigate the stability behavior of the flow in rectangular thermosiphon loops. A wide range of Lt/D ratio from 10 to 200 has been considered and the results show many different outcomes compared to previous 1D analytical calculations or stability theory. Moreover, we analyzed the flow in rectangular thermosiphon loops using Proper Orthogonal Decomposition (POD) and we observed that the cases without flow reversal are characterized by swirl modes typical of bent pipes and high-frequency oscillation of the related time coefficients obtained by Galerkin projection. However, the swirl mode was not observed in cases with flow reversals, these cases are characterized by symmetric flow field at 2nd POD mode and the similarity of low-frequency oscillation in the projection of POD modes.

physics.flu-dyn

An Energy-Based Lengthscale for Reduced Order Models of Turbulent Flows

In this paper, we propose a novel reduced order model (ROM) lengthscale that is constructed by using energy distribution arguments. The new energy-based ROM lengthscale is fundamentally different from the current ROM lengthscales, which are built by using dimensional arguments. To assess the novel, energy-based ROM lengthscale, we compare it with a standard, dimensionality-based ROM lengthscale in two fundamentally different types of models: (i) the mixing-length ROM (ML-ROM), which is a ROM closure model; and (ii) the evolve-filter-relax ROM (EFR-ROM), which is a regularized ROM. We test the four combinations (i.e., ML-ROM and EFR-ROM equipped with the energy-based and dimensionality-based lengthscales) in the numerical simulation of the turbulent channel flow at $Re_{\tau} = 395$. The numerical investigation yields the following conclusions: (i) The new energy-based ROM lengthscale is significantly (almost two orders of magnitude) larger than the standard dimensionality-based ROM lengthscale. As a result, the energy-based lengthscale yields more stable ML-ROMs and EFR-ROMs than the dimensionality-based lengthscale. (ii) The energy-based lengthscale displays the correct asymptotic behavior with respect to the ROM dimension, whereas the dimensionality-based lengthscale does not. (iii) The energy-based lengthscale yields ML-ROMs and (when significant filtering is effected) EFR-ROMs whose parameters are less sensitive (i.e., more robust) than the parameters of the ML-ROMs and EFR-ROMs based on the dimensionality-based lengthscale. The novel energy-based lengthscale could enable the development of better scale-aware ROM strategies for flow-specific applications and is expected to have long term applications in nuclear reactor thermal-hydraulics.

math.NA

Direct Numerical Simulation of Low and Unitary Prandtl Number Fluids in Reactor Downcomer Geometry

Buoyancy effect on low-flow condition convective heat transfer of non-conventional coolants, such as liquid metal and molten salts, is a crucial safety factor to advanced reactors under transient or accidental scenarios. The distinct heat transfer characteristics of non-unitary Prandtl fluids and the inherent complexity of the low-flow mixed convection phenomena requires the development of novel turbulent and heat transfer models that are adaptive to different spatiotemporal scales involved in the mixed convection heat transfer. In this work, direct numerical simulation of low-flow mixed convection is carried out at low-to-unitary Prandtl numbers that are of industrial interest. Time-averaged statistics, turbulent Prandtl number, as well as time signals are analyzed to investigate mixed convection phenomenon. From the time-averaged statistics, buoyant plume altered velocity boundary layer as well as the intensity of the fluctuation near both walls and channel centerline. Buoyancy effect also rendered different degree of convective heat transfer enhancement and impairment depends on Prandtl and Richardson number. Analysis of time series was conducted on the sodium mixed convection case to emphasize on the low-Pr mixed convection behavior at transition region. Resulting power spectra density and wavelet spectrogram suggests possible large convective structure in transition region. Future work will focus on providing broader data coverage on Pr-Re-Ri parameter space to facilitate more comprehensive analysis of mixed convection.

physics.flu-dyn

Direct Numerical Simulation of High Prandtl Number Fluid Flow in the Downcomer of an Advanced Reactor

The passive safety is a crucial feature of advanced nuclear reactor (Gen IV) design. During loss of power scenarios, the downcomer plays a crucial role. The fluid-flow behavior in the downcomer can involve forced to mixed to natural convection and characterizing the heat transfer for these changing regimes is a daunting challenge. High-resolution heat transfer numerical database can potentially support the development of precise and affordable reduced resolution heat transfer models. These models can be designed based on a multiscale hierarchy developed as part of the recently DOE-funded center of excellence for thermal-fluids applications in nuclear energy. In this paper, the downcomer is simplified to heated parallel plates and High Prandtl number fluid (FLiBe) is considered for all simulations. The calculations are performed for a wide range of Richardson number from 0 to 400 at 2 different FLiBe Prandtl numbers (12 and 24) which result in 40 simulated Cases in total. Time averaged and time series statistics as well as Nusselt number correlations are investigated to illuminate the mixed convection behaviors. The calculated database will be instrumental in understanding the flow behaviors in the downcomer. Ultimately, we aimed to evaluate existing heat transfer correlations and some modifications will be proposed for Cases where no satisfactory choice is available.

physics.flu-dyn

Direct Numerical Simulation of high Prandtl number fluids and supercritical carbon dioxide canonical flows using the spectral element method

The design of advanced nuclear reactors (Gen IV) involves an array of challenging fluid-flow issues that affect safety and performance. Currently, these problems are addressed in an ad-hoc manner at varying scales which are time-consuming and expensive. The creation of a high-resolution heat transfer numerical database has the potential to help develop to accurate and inexpensively reduced resolution heat transfer models. Such models can help address industrial-driven issues associated with the heat transfer behavior of advanced reactors. The models can be developed using the multiscale hierarchy developed as part of the recently DOE-funded center of excellence for thermal-fluids applications in nuclear energy. Ultimately this can lead to fast-running reliable models, thus accelerating the deployment of advanced reactors. In this paper, we performed a series of Direct Numerical Simulation using the spectral element codes Nek5000 and NekRS to investigate heat transfer in mixed convection conditions. First, we investigate the heat transfer of the flow in heated parallel plates for high Prandtl number fluids. The calculated database will eventually be used to evaluate existing heat transfer correlations and some modifications will be proposed for cases where no satisfactory choice is available. We have also investigated the heated transfer alteration phenomena in a straight heated tube for supercritical carbon dioxide (sCO2). The low-Mach-number approximation is used to decouple thermal and dynamic pressure, as pressure drop is negligible in this problem. The properties of sCO2 are calculated using multi-region polynomials. We observed that the heat transfer deterioration occurred in combination with the property changes of sCO2 and the depreciation of turbulence kinetic energy (TKE) for upward flow. Whereas, in downward flow, the heat transfer is enhanced thanks to the increase of TKE.

physics.flu-dyn