Conference Agenda

In a collaborative effort between Argonne National Laboratory and TerraPower, the high-fidelity computational fluid dynamics (CFD) code NekRS is being used to support the Natrium^® demonstration project. The overall aim of this effort is to use high-fidelity results to augment the available experimental data being used to validate the fast-running lower-fidelity tools used for reactor design. As part of this, NekRS has been used to simulate flow in a 37-pin wire-wrapped bundle using an LES turbulence model based on a high-pass filter. This replicates experiments conducted at MIT by Cheng. This study aims to corroborate the Cheng experimental flow split results via independent means, and validate the methodology used in NekRS for predictions of velocity in wire-wrapped assemblies. By validating NekRS for velocity predictions in wire-wrapped bundles, a firm basis for future work using the same methodology is established. Specifically, flow split at a Reynolds number of 16,170 is investigated and agreement is shown between the NekRS and experimental results to within experimental uncertainty. Details of the methodology will be discussed in the paper, including meshing, the turbulence model, convergence criteria and post-processing techniques. Additionally, the advantage of using a high-fidelity approach will be demonstrated by investigating flow phenomena which were not observable in the original experimental data, such as the velocity in the corner subchannels and the velocity skew across the assembly.

Hydrogen management during severe accidents at nuclear power plants has attracted attention as an important issue since the hydrogen explosion at the Fukushima Daiichi nuclear power plant accident in March 2011. In order to improve hydrogen management under severe accident conditions, the propagation of flames and the resulting loads on structures need to be predicted accurately. For this reason, the use of computational fluid dynamics is expected. Various benchmark experiments have been conducted, and turbulence models, turbulent combustion models, and chemical reaction models have been discussed. However, the uncertainties of each model have not been treated independently. Analysis with uncertainty quantification is necessary to promote efficient research activities through uncertainty-based prioritization and to reflect the latest findings in best practice guidelines. This study aims to establish a methodology for quantifying the chemical reaction uncertainty in turbulent premixed combustion CFD and performs the analyses on existing benchmark experiments. The uncertainties in the rate coefficients for the hydrogen combustion reaction were propagated through a one-dimensional flame propagation analysis to estimate the laminar flame speed uncertainty. Furthermore, the laminar flame speed uncertainty was propagated to a Reynolds-Averaged Navier-Stokes simulation using the turbulent flame speed closure (TFC) model to determine the mean and standard deviation of the maximum flame speed and maximum pressure.

Vapor condensation in the presence of a noncondensable gas is an important topic with practical applications in nuclear reactor safety. Passive Containment Cooling Systems (PCCS) involve shell-and-tube heat exchangers where steam is condensed inside tubes that are cooled by water pools.

Analytical models have been developed to estimate the heat removal of a tube condenser in such conditions. These models involve iterative and marching procedures, which is not warranted in fast running system codes. There is thus the need for direct correlations that provide accurate estimates of total condensation rates when the condenser wall temperature results from the interplay between the tube and shell sides.

A CFD model has been developed (Dehbi et al., 2013) and extensively validated under prescribed condenser wall temperature. In this investigation, we extend the validation to address conjugate heat transfer where both the shell and tube heat transfer are considered.

Two experiments are selected as validation databases, namely the Kuhn tube tests (1997), and the CONAN flat plate tests (2008). Both of these experiments involve well instrumented test sections that allow detailed information to be gathered, e.g. local wall/gas temperatures and heat fluxes.

Excellent agreement between the CFD predictions and experimental data is achieved, with heat flux deviations typically less than 5%. Since the CPU requirements are modest, the CFD model can thus be used in a parametric fashion to provide a numerical database from which easily implementable correlations can be developed using machine learning algorithms. This will be the object of a future investigation.

Phenomena involving complex multiphase gas-liquid flows, encompassing elements such as bubbles and free surface flows, are commonly encountered in various industrial processes, including nuclear applications. When it comes to Computational Fluid Dynamics (CFD) simulations, capturing the transition from low to high void fraction conditions presents a formidable challenge, primarily due to the escalating intricacies at the gas-liquid interface. For instance, gas volume fractions within the range where churn-turbulent and slug flows are prevalent are dominated by exceedingly deformable bubbles. In this intricate scenario, a generalized multiphase CFD modeling approach known as GENTOP stands out. GENTOP adopts the concept of a fully-resolved continuous gas phase, wherein this continuous gas phase encompasses all gas structures that are sufficiently large to be resolved within the computational mesh. However, it is important to note that for a typical user, delving into the complexities and technical nuances of setting up multiphase flow simulations can be quite challenging and laborious.

Turbulent mass transfer strongly influences flow-accelerated corrosion (FAC), a critical issue in designing liquid-metal-based nuclear reactors. Accurate simulation of FAC requires modeling scalar transport processes involving species with very low diffusivities, leading to flows characterized by high Schmidt numbers (Sc). Under such conditions, boundary layers become exceptionally thin, making Eulerian computational approaches prohibitively expensive due to the extensive near-wall mesh refinement required. In our previous research, we proposed a two-layer wall model capable of representing the effects of Schmidt and Reynolds numbers on scalar diffusivity. However, the original model relied heavily on numerical integration, thereby increasing computational demands. To address this, we present a surrogate formulation with explicit integration, reducing computational complexity and simplifying integration into computational fluid dynamics (CFD) codes. This study extends validation efforts to challenging high-Sc-number scenarios involving orifice plate and slot flows under strongly non-equilibrium conditions. Simulations were conducted using the Abe-Kondoh-Nagano (AKN) low-Re k–ε turbulence model. Results confirm that our surrogate two-layer model maintains excellent accuracy in predicting peak near-wall mass transfer without the necessity of extensive mesh refinement (with first-wall grid spacing maintained at y⁺ above 1). Moreover, the model demonstrates improved predictions compared to wall-resolved approach from the literature, especially in capturing non-equilibrium effects downstream of flow disturbances. These findings illustrate that the developed surrogate two-layer model provides both computational efficiency and enhanced accuracy, making it highly suitable for engineering applications involving high-Schmidt-number mass transfer phenomena and FAC predictions.

The Compact Steam Generator (CSG) could play a crucial role in the development of Small Modular Reactors, particularly for the Integral-Pressurized Water Reactor (iPWR), which is gaining significant attention due to its potential to provide safe, reliable, and cost-effective nuclear energy. The Printed Circuit Heat Exchanger (PCHE) is a promising candidate technology that could meet the requirements of the CSG. This study examines the capabilities of existing Computational Fluid Dynamics models for the Printed Circuit Heat Exchanger, considering both single-phase and multiphase conditions, with a focus on the mixture-multiphase approach using the Rohsenow boiling model. The steam generator conditions involve boiling heat transfer, transitioning from subcooled liquid to high-quality steam. This results in a high gradient of mixture density and flow acceleration, which may pose challenges for the CFD solver. This study will discuss these challenges and assess the employed methodology. The results are validated against experimental data from the Georgia Institute of Technology, which conducted experiments on a semicircular channel (≈ 2 mm) PCHE under a wide range of conditions. The results demonstrate good agreement between the simulation and experimental data for both single-phase and multiphase flows across a broad range of conditions, despite the Rohsenow model being developed for pool boiling. Furthermore, the Rohsenow model tends to overpredict heat transfer; therefore, additional calibration of the model may lead to slight improvements in predicting a wide range of flow boiling conditions.

To evaluate the impact of uncertain input parameters on numerical models, Sensitivity Analysis (SA) is an invaluable tool. It supports the process of quantifying uncertainties in the outputs of numerical simulators used to model and predict physical phenomena. This helps in understanding how these uncertainties influence the model outputs. Among the methods of SA, the one based on estimating HSIC (Hilbert-Schmidt Independence Criterion) indices is particularly interesting. This approach is especially useful when each run of the simulator is CPU-time expensive, as HSIC indices can be well-estimated with fewer than a hundred simulations.
This paper presents the results of a HSIC-based Global SA (GSA), applied to a multi-physics simulator (named CORPUS) modeling a load rejection transient in a Pressurized Water Reactor. Three uncertain input variables were considered: control rod worth, moderator temperature coefficient and fuel conductivity. Two CORPUS outputs, which are time-dependent variables, were selected as quantities of interest: the core power and the axial offset. From a Monte-Carlo sample of one hundred CORPUS simulations, a GSA based on a functional Principal Component Analysis (PCA) combined with the estimation of HSIC indices on the PCA coefficients is performed for each of the two outputs.
The analysis reveals that the most important uncertainty source for the core power is control rod worth, followed by fuel conductivity and then by moderator temperature coefficient. The most important uncertainty source for the axial offset is again control rod worth followed by moderator temperature coefficient, while the impact of fuel conductivity is negligible.

As part of research to ensure pressurized water reactor safety, Thermo-hydraulic (TH) and neutron kinetic tools are deployed to predict scenarios of intense local boiling and radial void fraction within innovative fuel assemblies, which can strongly influence reactor power. Current validation of the TH system tools is limited under these particular conditions (pressure around 70 bar, high void fraction), and given the restricted availability of experimental data in the literature, a comparison with CFD simulations is exploited to support the system scale.

The ultimate goal is to validate the porous 3D module of the system code CATHARE3 (C3) on transverse two-phase flows between parallel sub-channels. The work starts with the validation of a multi-scale post-processing method (from CFD tool neptune_cfd to C3) on a single channel experimental test case from the PSBT benchmark. It is then extended to a two sub-channels geometry from “Rowe and Angle” 1967 experiment, which focuses on two-phase cross-flows.

The multi-scale post-processing method has proven to be a valuable tool for comparing results between the different codes’ scales. Numerical results from C3 and neptune_cfd are in good agreement with experimental data from PSBT experiment. However, results from Rowe and Angle's experiment show that the C3 turbulence model overestimates turbulent viscosity, resulting in inaccurate two-phase cross-flow predictions. Alternative models are tested to improve C3 code prediction.

Propagating uncertainties in nuclear thermal-hydraulics simulations is challenged by the anal-
yses in complex domains under extreme operating conditions and multi-physics. The compu-
tational cost becomes particularly high for liquid metal cooled reactors, for which the similar
thermal conductivity of fluids and solid parts require a Conjugate Heat Transfer (CHT) ap-
proach.
In the present work, we applied a methodology based on surrogate modeling and data of
multiple fidelities to propagate these uncertainties to the maximum cladding temperature. The
correlation between the data collected for fluid and solid regions is exploited to retrieve uncer-
tainty estimates for the cladding temperature with a limited number of CHT simulations. The
computed probability distributions show the impact of deformation on the excess temperature
and highlight the different response to deformed bulk, edge and corner pins.

Accurate estimation of mesh related numerical errors and the associated uncertainties is critical in computational fluid dynamics (CFD) simulations to ensure the credibility of the results. Traditional Richardson extrapolation-based approaches often exhibit limitations when applied to turbulent flow regimes. In real-world CFD applications, factors such as turbulence models, complex geometries, and algorithm limiters can result in nonlinear responses during numerical convergence studies, leading to unphysically large uncertainty bounds. Notably, these uncertainties stem from the error estimation approach itself, rather than the CFD solver, and hinder the consistent application of CFD to complex reactor simulations. In this work, we present a robust stochastic framework for estimating mesh related uncertainty in unsteady turbulent flow simulations. The framework starts with quantifying the potential numerical errors in local turbulence characteristics using the least-squares method; the errors are then propagated into the system through the physics-based stochastic field modeling. The framework removes the deficiencies of the conventional approaches and properly accounts for the intricate interactions between numerical and model error. The impact of numerical uncertainties on turbulence predictions and, consequently, on the overall flow field predictions of unsteady flow can be well described by propagating the local turbulence characteristics. To demonstrate the efficacy of the proposed approach, an unsteady mixed-convection problem is presented. The results show that, compared to the conventional approach, the proposed framework is robust in estimating potential numerical error. Moreover, the framework can identify the error caused by poor spatial/temporal resolutions and alert the user to the quality of the numerical model.

This study explores the capabilities of the Generalized Turbulent Flame Closure approach, recently presented by Karanam and Verma, for scalable Deflagration-to-Detonation Transition (DDT) modeling in hydrogen-air mixtures. The TFC-DDT approach introduces key simplifications that enable DDT simulations on underresolved meshes, which could make it suitable for large-scale applications, including those relevant for nuclear safety. This work offers re-implementation of the TFC-DDT model within the nuclear-focused open-source combustion framework flameFoam to conduct its independent validation.

The validation focuses on the model’s ability to capture flame acceleration, detonation onset, and shock behavior under varying conditions, while also examining grid resolution requirements and sensitivity to numerical parameters. This investigation aims to further understand the TFC-DDT model's potential for large-scale detonation applications in safety and industrial contexts, contributing to the advancement of simplified, computationally efficient DDT modeling techniques.

Given the lack of validated systems level open source codes for coupled natural circulation, a model for coupled natural circulation in the Compact Integral Effects Test (CIET) was built using the Open Source Thermo-hydraulic Uniphase Solver for Advection and Convection in Salt Flows (TUAS). This model was validated using experimental data of natural circulation mass flowrate within CIET at various prevailing boundary conditions. The resulting CIET model built in TUAS agreed well with the experimental data as the discrepancy between the TUAS model and experimental data was comparable to the discrepancy between the SAM model and experimental data. Moreover, the TUAS model of CIET was able to run in real-time on a personal computer due to simplifications such as the Boussinesq approximation and the fact that it was built using the Rust programming language which has execution speed comparable to Fortran and C++. These results show that TUAS is potentially useful for systems level code analysis, digital twin applications as well as real-time reactor simulators. Its open source license also contributes to repeatability of results and the potential to expand it to many applications. This could greatly contribute to the molten salt thermal hydraulics community as an additional tool for systems level reactor analysis.

Tritium extraction is a universal challenge in advancing all types of nuclear power: fission plants must dispose of, and fusion plants must fabricate fuel from it. Gas-liquid contactors (GLCs) which leverage two-phase dynamic mixing to supercharge the mass transfer of tritium are an untapped concept in the field. In theory GLCs should be capable of achieving more surface area than static liquid-solid interface extractors with sufficiently small & many bubbles, but bubble coalescence has been a major challenge. By building on Volume of Fluid (VoF) multiphase simulation support in the computational fluid dynamics tool OpenFOAM, we developed an extensible Continuum Species Transport (CST) solver for the loosely coupled concentration field of a dilute miscible species. This establishes a new framework for the design and pre-experimental evaluation of novel GLC concepts in mass transfer applications for molten salt. To demonstrate the utility of our new solver, a simple evaluation of a bubble column with one large Ar inlet against four small Ar inlets is performed, showing a straightforward correlation between bubble size and tritium extraction efficiency.

Molten salt reactors (MSRs) have gained much interest in the nuclear community over the past few years, and efforts are currently being made in the design and deployment of a molten salt research reactor (MSRR) at Abilene Christian University. Multiple experimental salt loops are being designed to test various aspects of MSRs and understand the fluid dynamics of molten salts at a higher level. One such experiment is a bubble flow salt loop built at Texas A&M. High fidelity models of this experiment have been constructed, utilizing the computational fluid dynamics (CFD) modules of the MOOSE software. These CFD models, being at a high fidelity, may provide important information and a more wholistic view of the mechanics of molten salts, and may be included as a digital twin (DT) component in the experimental loop, as well as provide useful information to the MSRR. However, these CFD models require computationally expensive runs to provide reasonable and usable data. To combat this, a reduced order model (ROM) algorithm will be developed for these CFD models, solving these high fidelity and high dimensional problems in a significantly lower dimensional latent space, reducing cost with acceptable losses in accuracy. For these models, various ROM algorithms are created, using methods such as proper orthogonal decomposition (POD), neural networks (NN), and convolutional neural networks (CNN). These algorithms are then tested in an offline mode, comparing the forward propagation of the lower dimensional problem against the propagation of the full dimensional model.

The Stable Salt Reactor (SSR) integrates features of molten salt reactor technology with conventional light water reactor fuel assembly designs to achieve enhanced safety and economic benefits. Utilizing a fast reactor configuration, the SSR employs recycled nuclear waste as fuel, contained within salt-filled fuel pins submerged in a liquid salt coolant. Effective heat transfer between the molten fuel salt and coolant salt is critical to the reactor's core safety and operational reliability.

This study investigates conjugate heat transfer (CHT) processes in the SSR's narrow salt-filled fuel pins using the high-fidelity spectral element computational fluid dynamics (CFD) code, NekRS. The analysis encompasses internal natural convection within the molten fuel salt and external forced convection in the liquid salt coolant under both normal and transient conditions. Parametric studies are conducted to assess the influence of reactor power and coolant flow rates on heat transfer performance. The resulting data are leveraged to develop and validate heat transfer models for integration into the SAM system code, facilitating efficient transient safety analyses. The findings of this work refine safety system models, enhance the predictive accuracy of SSR core designs, and quantify uncertainties in molten salt CHT simulations. This study highlights the critical role of advanced CFD technologies in expediting the engineering design and licensing of next-generation nuclear reactors like the SSR.

Molten Salt Reactor (MSR) is currently one of the most promising Generation IV reactors, actively being developed internationally due to its high economic efficiency and safety. While evaluating the economic feasibility of developing MSR is important, assessment of various accident scenarios is also required for safety evaluation. One of the most anticipated scenarios in MSR is a salt spill accident caused by the crack or rupture of reactor pipes and the reactor vessel. In such cases, it is necessary to effectively cool the spilled molten salt and contain it within the desired location such as drain tank. To design an efficient molten salt transport structure, a detailed analysis of the molten salt behavior is essential. Furthermore, there is a possibility of releasing fission products into the atmosphere in the form of aerosols from the molten salt, for which boundary conditions can be provided. Lagrangian-based Computational Fluid Dynamics (CFD) calculations are considered a more advantageous numerical method for analyzing solidification behavior compared to Eulerian CFD methods. This is due to its meshless analysis characteristics, which allow free changes of boundaries between fluid and wall according to phase changes. In this study, the Moving Particle Semi-implicit (MPS) method was used to analyze salt spill behavior in MSR. To accurately simulate behavior accompanied by solidification, a model was developed to account for heat transfer and wall adhesion based on phase changes. A comparative analysis was conducted with results from other numerical methods, including Eulerian-based analysis.

Two-phase flow in Molten Salt Reactors (MSRs) is important as it impacts reactivity evolution, reactor transient response, and the removal of species dissolved in the molten salt through gas phase transfer. Therefore, accurately predicting the gas distribution and the associated liquid-gas interface area in MSRs is essential for their design and operation.

Recently, we integrated two new models into Idaho National Laboratory (INL)’s Multiphysics Object-Oriented Simulation Environment (MOOSE): a multi-D generalization of a mixture drift-flux model and a Euler-Euler model. The Euler-Euler model offers higher fidelity, while the mixture drift-flux model provides greater computational efficiency, which is typically preferred for modeling reactor transients. However, the mixture model's accuracy in capturing void distribution and interfacial area in MSRs still needs to be assessed.

This article begins with a description of the mathematical framework for the two-phase models implemented in MOOSE. It then presents validation of these models against relevant experimental data. Finally, both models are applied to the Molten Salt Reactor Experiment case study, analyzing various operational conditions such as different rates of fission product volatilization and diverse cover gas entrainment scenarios at the reactor pump. The article concludes by assessing the suitability of both models for capturing the two-phase flow dynamics critical to MSR operations.

Argonne is participating in the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop (NACIE)”. The NACIE loop includes a fuel pin simulator section which is a hexagonal array of 19 wire-wrapped, electrically heated pins and uses lead-bismuth eutectic as a working fluid. Argonne’s work on the CRP includes CFD simulations with the NekRS and Cardinal codes. Both LES and RANS turbulence models were used in NekRS coupled via Cardinal to a solid conduction model to account for conjugate heat transfer. Via comparisons against experimental measurements from the NACIE tests, these benchmark simulations are being performed to expand the validation basis of these codes. The objective of this paper is to present recent progress on NACIE test simulations which cover both forced and mixed convection conditions with uniform and skewed heating profiles. Results from simulations will be compared to available experimental data.

CFD becomes more and more important in nuclear reactor safety considerations. For multiphase flows in the related medium and large scales the Euler-Euler approach is most frequently used and often the only feasible one. In many flow situations, the involved interfaces cover a wide range of scales leading to different coexisting morphologies. Established simulation methods differ for the different interfacial scales. Large interfaces are represented in a resolved manner usually basing on the one fluid approach, e.g. Volume of Fluid (VOF) or Level Set. Unresolved (dispersed) flows are modelled using the two- or multi-fluid approach. A simulation method that requires less knowledge about the flow in advance would be desirable and should allow describing both interfacial structures – resolved and unresolved – in a single computational domain.

The morphology adaptive multifield two-fluid model MultiMorph, which is developed at HZDR based on the software from the OpenFOAM Foundation, is able to handle unresolved and resolved interfacial structures coexisting in the computational domain with the same set of equations. An interfacial drag formulation for large interfacial structures is used to describe them in a VOF-like manner, while the usual closure models are applied for the unresolved phases. In addition, MultiMorph allows to simulate transitions between the morphologies. This concerns both empirical transitions such as entrainment and detrainment as well as transitions resulting from a change in the size of the numerical mesh within the domain. The basic framework including the handling of transitions between the morphologies will be presented.

Blockage in a reactor fuel assembly is considered to be one of the most important accidents that should be analyzed in detail. A variety of factors can contribute to the occurrence of such an accident, among which are fuel element bending or local deformation and swelling of the cladding. The reduction of coolant flow area can cause local heat transfer deterioration and temperature augmentation, which can further lead to dry-out and possible loss of fuel assembly integrity. It is challenging to evaluate the consequence of flow blockage accident due to lack of knowledge about local flow parameters as well as their response mechanism, especially when two-phase flows are concerned. Moreover, due to negligible influence on global mass flow, it is difficult to detect the accident through protection system. Owing to the availability of advanced computer systems, simulation using either system or CFD codes has become an important tool in assisting the analysis of these local phenomena and evaluation of their impact on the safe operation of nuclear reactors. This study presents a CFD study of three vertical pipe bubbly flow cases, one empty pipe, one with a ring obstacle and one with a baffle obstacle, and both obstacles block a half of the flow area. The focus is put on analyzing the effect of blockage on the velocity and turbulence field as well as the development of bubble size. Different turbulence models and mechanisms leading to bubble coalescence and breakup are discussed and evaluated with the aid of experimental data.

During severe nuclear accidents, hydrogen generated in the reactor core can mix with air to form potentially explosive mixtures. The severity of such explosions depends on the mixture composition and the combustion regime. In highly turbulent conditions, combustion is significantly accelerated, and simulations have shown that heat loss has minimal impact on flame propagation due to its slower rate relative to combustion. However, in the case of slower deflagration, heat loss is expected to play a more significant role in flame evolution.

This study focuses on modeling the effects of conductive and radiative heat losses in a slow hydrogen-air-steam combustion scenario during the HD-22 experiment at the THAI experimental facility. Unsteady Reynolds-Averaged Navier-Stokes (RANS) simulations were conducted using computational fluid dynamics software OpenFOAM software and combustion model flameFoam. Heat loss was modeled through conductive heat transfer to an isothermal wall and the P1 model for radiative transfer.

Simulation results showed good agreement with experimental data, indicating that including heat loss mechanisms slightly delayed the completion of combustion and slighlty reduced the maximum overpressure, as well as slowed down the vertical flame propagation. Results show that the conductive and radiative heat loss contribute similarly to the total heat loss, emphasizing radiative heat loss importance in modeling slow combustion.

Overall, the study highlights the critical role of heat loss, particularly radiative heat transfer, in accurately simulating slow hydrogen explosions.

The pressure drop of PWR fuel assemblies has an importance in the core itself (global flowrate, flow distribution, hydrodynamic forces) but also in the storage pools and in the different cells or casks in which it may be stored. The behavior in nominal conditions is well known and has been largely experimentally and numerically investigated, however the characteristics at much lower Reynolds numbers are less studied.

The first objective of this paper is to discuss the adequate turbulence models for those low turbulence situations. Available experiments provide useful data of pressure drops and velocity of different elements in a large range of Reynolds numbers (device made of a typical fuel bundle (17x8) with 4 mixing and supporting grids, at full scale). CFD computations with RANS, URANS, Detached Eddy Simulations and Large Eddy Simulation turbulence models are performed and compared with measurements. The modeling of the wall friction is also discussed. The computational meshes are obtained from an automatic process and are based on polyhedrons.

The second objective is the influence of the domain modeled in such quasi-periodic configurations and the associated boundary conditions. The different periodicity conditions are particularly investigated on 2x2 simplified models, with the transverse effect of mixing vanes.

The commercial Ansys-Fluent and Star-CCM+ codes are both used and conclusions are drawn on the two objectives mentioned above.

Finally, this paper concludes on the recommended setup in order to provide industrial values of pressure drops of the components of the fuel assembly in the different off-reactor situations.

The Forschungs-Neutronenquelle Heinz Meier-Leibnitz (FRM II) is actively contributing to global efforts to reduce the use of Highly Enriched Uranium (HEU) in civilian nuclear applications. A key step in this initiative is converting the current fuel system to a high-density Low Enriched Uranium (LEU) fuel. In 2023, a feasibility study demonstrated the scientific viability of converting the FRM II to U-10Mo LEU, while maintaining FRM II scientific performance. Due to the changed plate design, many LEU designs exhibit lower pressure drops through the fuel element. To minimize the impact to the FRM II reactor, we aim to have a similar pressure drop across the fuel element in the forward flow direction than for today’s HEU core and a lower pressure drop required in reverse to mitigate impacts during transient scenarios. To meet these demands, the previously published design incorporated a flow restrictor downstream of the fuel element to increase the pressure drop. In the current study, an alternative approach is explored by thickening the cladding at the end of the fuel plates. Various cladding thicknesses are evaluated, and the resulting pressure drops are computed using Computational Fluid Dynamics (CFD). Additionally, the influence of rounded fuel plate caps is quantified.

Nuclear thermal hydraulic analyses generally employ system and subchannel codes to examine the safety and performance characteristics of a given reactor system. More recently, 3D CFD modelling has been widely used to produce higher fidelity analysis but such methods can mainly be used for local phenomena. This research seeks to develop a coarse-grid CFD model using a subchannel approach to compute wall effects so as to create a computationally cost-effective model for the two-phase flow dynamics within the nuclear reactor core. The subchannel CFD (SubChCFD) technique previously developed for single-phase flow at the University of Sheffield is first implemented in OpenFOAM^TM CFD solver. The 5X5 bare rod bundle case from the NESTOR experiment is used for validation. Model predictions are compared well against experimental data. This single phase model has then been extended to the homogeneous equilibrium two-phase flow model concept and models for the wall effects (including wall boiling) and phase crossflows are being developed. In the full paper, the OpenFOAM^TM implementation and validation for both single and two-phase boiling flows will be discussed and evaluated.

The presence of spacer grids and mixing vanes in a typical Pressurized Water Reactor Fuel Assembly leads to significant turbulence in the coolant sub-channels, which plays a very important role to enhance the heat transfer performance of the fuel assembly. In this preliminary work, OpenFOAM is used to build a single coolant channel model with fuel rods, spacer grid, and mixing vanes, with periodic boundary conditions, to study the complex fluid flow behaviour induced by the mixing vanes and spacer grid. The flow field results obtained from OpenFOAM with the Reynolds Averaged Navier-Stokes (RANS) turbulence models are found to match well with experimental results from the MATIS-H benchmark. In addition, multiple hyperparameters are varied to study their effects on computational resources required and stability, and accuracy of results.

The CFD task group of the OECD Nuclear Energy Agency (NEA) Working Group on Analysis and Management of Accidents (WGAMA) drives collaborative activities in Computational Fluid Dynamics for Nuclear safety studies since the early 2000s. This paper presents recent achievements and the current major objectives of the group, illustrated by examples of code benchmarks and reviews of reference documents such as the updated Best Practice Guidelines.

A recently published Technical Opinion Paper identifies main challenges towards the perspective use of CFD in safety studies. Several ongoing and future activities identified and initiated will be discussed. In the qualification process of scientific computational tools , the quantification of uncertainties (UQ) plays a major role in the estimation of the trust of a given evaluation. As far as CFD is concerned, several specificities, mostly induced by the computational cost have been recently analyzed within an expert group composed of CFD, UQ and data science specialists. This paper shares the outcomes and potential future activities on the topic.

In the context of the qualification process, identification of suitable application-oriented validation data is an important step. WGAMA is currently considering an important task concerning the update and extension of the CSNI Code Validation Matrix (CCVM) for reactor coolant system and containment thermal-hydraulic phenomena of current and advanced water-cooled reactor including SMR. In the field of CFD, a connected activity is foreseen on the identification and analysis of validation data for specific safety applications. The needs and way forward will be discussed within this paper.

To meet the requirements of three-dimensional numerical simulations for the flow
and heat transfer phenomena in nuclear reactors, this article introduces WINGS-CFD （Workbench of Intelligent Nuclear reactor desiGn and Simulation-Computational Fluid Dynamics）, an Computational Fluid Dynamics (CFD) code for nuclear reactors developed by Nuclear Power Institute of China, which is designed based on the object-oriented and hierarchical architecture principles with a high degree of extension. This article comprehensively outlines the overall design concepts of WINGS-CFD, covering aspects such as theoretical models, numerical discretization methods, and code architecture. Several numerical calculations for typical reactor scenarios involving flow and heat transfer conditions are conducted using WINGS-CFD and the commercial CFD software Fluent, and the results demonstrate that the computational accuracy of both codes is comparable. Furthermore, WINGS-CFD is capable of supporting large-scale numerical simulations of neutron transport and flow heat transfer with billions of cells and it exhibits superior parallel performance. This WINGS-CFD code provides an autonomous numerical technique for the fluid dynamics and heat transfer analysis of reactor systems.

In the present work, a novel methodology for the numerical simulation of single-phase flow in rod bundles, within a coarse-mesh context, is presented. The proposed approach aims to fill the gap between standard CFD and subchannel modeling, targeting a stronger balance between the accuracy of the numerical solutions and a reduced computational effort. For this purpose, different numerical techniques were designed and combined in custom applications within the OpenFOAM framework. In the first place, wall models for the turbulent quantities that rely on the use of empirical correlations were implemented. The use of these models reduces the traditional restrictions on the mesh refinement at the walls and consequently introduces significant gains in computational efficiency for a similar level of accuracy when compared to standard simulations. On the other hand, special numerical schemes were implemented to include the capability of handling piecewise linear pressure distributions in a finite volume context, i.e., discrete changes in pressure across selected locations that mimic the behavior of unresolved, localized geometrical scales. This approximation is used to model spacer grid effects without explicitly including the spacer grid in the geometry. Additionally, new numerical techniques that allow the use of different meshes for each equation are explored. The combination of the proposed coarse-mesh approximations is validated against high-resolution numerical simulations and experimental data.

With the advancement of nuclear technology, the demand for precise simulation of flow fields in reactor rod bundles has significantly increased. Traditional low-order eddy viscosity models often lack the accuracy needed for complex subchannel flow fields, while high-order numerical methods like Large Eddy Simulation (LES) provide high-fidelity flow structures but are computationally expensive, limiting their practical application. To address the challenge, this study introduces a Condition Adaptive Eddy Viscosity Turbulent Model (CAEVTM) tailored for specific subchannel conditions in rod bundle with Mixing Vane Grids(MVGs). The goal is to achieve a balance between high accuracy and low computational cost by integrating high-order numerical simulations with machine learning techniques. The research begins with performing scale-resolving simulation to obtain high-fidelity flow structures. Key physical quantities sensitive to subtle flow changes are then selected to ensure the model's responsiveness. Subsequently, gradient-based techniques are employed to calculate the sensitivity of each grid point, and a optimization method is used to optimize the correction coefficients, enhancing the model's accuracy. A neural network is then trained to map the operational conditions to the correction coefficients efficiently. Finally, the trained neural network is incorporated into the secondary eddy viscosity model to construct the CAEVTM. Results demonstrate that CAEVTM significantly improves the accuracy of flow field predictions in specific subchannels while greatly reducing computational costs. The proposed CAEVTM successfully combines high-order numerical simulations with machine learning techniques to achieve efficient and accurate simulations of flow fields in specific grid subchannels.

Thermal radiation plays a crucial role in heat transfer for next-generation nuclear reactors due to the high operating temperatures and reliance on natural convection. In our previous work, we successfully implemented the P1 approximation for thermal radiation in Nek5000/NekRS, a CFD code based on the Spectral Element Method. The implementation was verified against both numerical data and analytical solutions. However, that work focused solely on the fluid domain.

In this paper, we extend the P1 model to the solid domain by integrating it with the Conjugate Heat Transfer model in Nek5000/NekRS. As before, we validate our implementation with reference numerical solution on simple geometries. Then, we apply this approach to the pebble bed case with 1,568 pebbles under salt flow cooling, which was introduced in our previous work, now including the solid domain of pebbles. To ensure accuracy, we conducted parallel simulations using OpenFOAM for comparison.

The presence of a large-scale interface between the two phases presents a modeling challenge related to turbulence in a sense that at a large-scale interface the lighter phase sees the heavier phase like a solid wall. In literature, a widely used strategy is to add a damping source term to the turbulence dissipation equation. These source-term based approaches require that a parameter be tuned based on grid and/or problem. In the present work, a novel approach is proposed where the use of solid wall-function based turbulence treatment is done within the framework of LSI model implemented in Simcenter STAR-CCM+. This is achieved by using a large-scale interface toolkit, which provides the location of the large-scale interface as well as the cell-center to large-scale interface distance; enabling creation of a stencil around the large-interface cell to apply wall-function base damping treatment.

The effectiveness of the new wall-function type interface turbulence damping method is demonstrated in this work though a turbulent air-water co-current stratified flow, studied experimentally by Fabre et al., (1987). The wall-function type treatment shows minimal dependence on grid refinement as well as on the cell aspect ratio when comparison is performed for the full developed velocity profile against the experiment.This is not the case for source term based approach. The observations made using the velocity profile are corroborated for the pressure drop values as well. In addition, the novel approach is shown to not need any problem-based or grid-based tuning.

Two-phase flows are of great interest to the nuclear power community, both in normal operation for coolant systems, and in accident scenarios. Computational multiphase fluid dynamics (CMFD) simulations are a promising tool for detailed analysis of these systems. However, since most CMFD models are based on experiments and analysis of vertical two-phase flows, significant limitations are evident when horizontal bubbly two-phase flows are considered, which have entirely different hydrodynamics arising from the large density difference between phases. In this work, a new experimental database of relative velocity for horizontal bubbly flows is established utilizing the existing 25.4 mm test facility at Purdue University. Local void fraction, gas velocity, and bubble diameter are measured with four-sensor conductivity probes, while local liquid velocity is measured with a Pitot-static probe. This data is used to evaluate the existing CMFD models in ANSYS CFX, demonstrating the problems with the existing models, especially with predicting relative velocity. A recently proposed relative velocity model is implemented in CFX, which improves both the relative velocity and void fraction prediction. Qualitatively, the void fraction location and shape are greatly improved, peaking further away from the wall with an elliptical shape that better agrees with the experimental data. The CMFD predicted area-averaged void fraction matches the experimental data about 10% better with the new model, while the relative velocity is improved by 30%.

Turbulence induced vibration of fuel rods can lead to mechanical wear, which can be responsible for safety issues and significant maintenance costs in Nuclear Power Plants (NPPs). In the context of the “Gathering expertise On Vibration ImpaKt In Nuclear power Generation” (GO-VIKING) European project, fluid structure interaction (FSI) methods are developed and assessed on simpler geometries. Hereby, two-way coupling simulation, of a cantilever rod under a turbulent axial flow have been performed at Reynolds number 21 200. Computational fluid dynamics (CFD) with finite volume method and wall-resolved large eddy simulation (WR-LES) is used for the flow, along with an in-house Euler-Bernoulli beam equation solver for the rod. Free vibrations tests have been performed for the calibration of the mechanical damping of the beam and for validation purposes. The beam motion is accounted for using an Arbitrary Lagrangian Eulerian (ALE) method. The meshes are made of 144 million hexahedra for the flow and 3800 grid points for the beam, respectively. The beam motion and the flow statistics are both investigated; comparisons are drawn with experimental results from the literature. The theoretical and experimental natural frequencies are recovered by the simulation. The R.M.S of the tip displacement of the beam is lower in the simulation than in the experiment. However, this may be due to both the experimental uncertainties at this Reynolds number and a misalignment of the rod in the experiment. Apart from the dissymmetry due to the misalignment, there is quite a good agreement regarding the flow statistics.

Natural convective heat transfer is a prominent research topic in the nuclear energy field and is crucial for the design of passive safety systems. To study the impact of compressibility-induced non-Oberbeck-Boussinesq (NOB-II) effects on natural convection, we conduct lattice Boltzmann simulations of square cavity natural convention in a perfect gas, incorporating the multiple-relaxation-time force model and pseudopotential force. The findings indicate that for a given Rayleigh number (Ra), the Nusselt number (Nu) increases as NOB-II effects strengthen and the Reynolds number (Re) decreases as these effects intensify. This implies that NOB-II effects lead to heat transfer enhancement and convection suppression. The underlying mechanism is as follows (taking the hot fluid as a representative case): under NOB-II conditions, the compression work term absorbs heat from the hot fluid near the central region of the hot wall, resulting in a steeper temperature gradient and a thinner temperature boundary layer near the hot wall. Consequently, the local Nusselt number increases and overall heat transfer is enhanced. Simultaneously, the reduction in the thickness of the temperature boundary layer causes a decrease in the buoyancy difference, ultimately leading to convection suppression. Furthermore, new scaling laws of Nu-Ra and Re-Ra considering NOB-II effects are proposed, with an average error of less than 5%. This study deepens the understanding of natural convection and offers theoretical support for the thermal-hydraulic and safety analyses of advanced reactors.

This study presents a computational fluid dynamics (CFD) methodology for simulating multi-regime two-phase flow in an in-line tube bundle. The approach is based on a mixture multiphase model, incorporating scale separation between large interfaces resolved according to the mesh size and the dispersed phase modeled using a three-dimensional slip velocity to account for kinematic disequilibrium between phases. The primary objective was to validate this methodology by comparing CFD results with experimental data. The quantities of interest are the Power Spectrum Density of the forces applied in a tube and the local distribution of the void fraction around an instrumented tube. Analysis shows good agreement between CFD results and measurements regarding the quantities of interest investigated. This work demonstrates the consistency and the reliability of the methodology for two different mixtures: Water/Air and fluids simulating water/steam. In addition, this approach reduces the computational complexity in comparison to two-fluid models while maintaining a good accuracy in predicting two phase flow topology in the tube bundles.

This work represents a significant advancement towards developing a one-fluid formulation methodology for simulating complex multi-regime two-phase flows in tube bundles, with particular emphasis on studying fluid-structure interaction (FSI) in the U-bend region of steam generators.

The OpenSTREAM computational environment is a new open-source platform designed to facilitate efficient and collaborative development and validation of one-dimensional, multi-field, two-phase flow simulation models across research institutions. It includes several simulation frameworks: a mixture model, a two-fluid model, a three-field model, and an advanced four-field model of annular two-phase flow. The current implementation supports single-component, incompressible, steady-state, and transient boiling two-phase flows in single straight channels under reasonable simplifying assumptions. The two-fluid model solves a six-equation system governing mass, momentum, and energy conservation for each phase, capturing hydrodynamic and thermal non-equilibrium effects. The three-field model follows a classical framework (vapor, drops and film) for annular two-phase flow, while the advanced four-field model explicitly represents both the base liquid film and dispersed disturbance waves as separate fields. In all solvers, field interactions and wall closure models have been implemented either from well validated models from the literature or from simple considerations, providing a foundation for future collaborative improvements. Simulations of a representative boiling water two-phase flow case using all simulation frameworks show consistent and reasonable predictions. A comparison with the TRACE system code demonstrates that the implemented two-fluid solver produces reliable and consistent results. Finaly, the validation exercises from the original four-field model development are reproduced. OpenSTREAM, along with its validation and application database, will soon be publicly available on dedicated GitHub repositories under the permissive MIT license.

Since innovative reactors' flow conditions and geometries may differ from those of conventional reactors, the applicability of the models and correlations used in the design works should be appropriately checked. In the design phase, there is a high possibility that the flow conditions and geometries will be changed. Therefore, applying detailed numerical simulation is expected to achieve efficient design works. In nuclear reactors, two-phase flow will appear in many situations. Confirming the applicability of models and correlations for two-phase flow conditions is important. The detailed two-phase flow simulation codes must be useful to confirm the applicability of two-phase flow models and correlations. However, there is no established methodology to properly validate detailed two-phase flow simulations. We have, therefore, started the research project to develop a methodology for validating detailed two-phase flow simulation codes. In this project, we have developed two-phase flow measurement technologies to obtain detailed information on the gas-liquid interface and two-phase flow database by using developed measurement technologies and performing detailed two-phase flow simulations for the developed two-phase flow database. We will compare detailed two-phase flow simulation results and the two-phase flow database and discuss the proper methodology with the reactor manufacturer. Finally, we will investigate the validation process of a detailed two-phase flow simulation code. In this presentation, we will show the outline of this project and future plans.

Two-phase liquid-gas flow has a wide variety of applications in the nuclear industry, including active thermal control systems, steam generators, and nuclear reactors. In order to model and predict the pressure drop and flow regimes in a reactor core, the void fraction must be accurately predicted. This paper presents a new mathematical model that can accurately predict the two-phase void fraction requiring only knowledge of the geometry of the channel, liquid and vapor mass flow rates, and properties of the working fluid. The predicted void fraction is validated by void fraction data collected using an in-house capacitance sensor and a unique vertical, air-water flow calibration loop. Compared to measured void fraction data, the new mathematical model has a better performance than commonly used models such as Lockhart Martinelli model, Wheeler model, Chen model and homogeneous model.

Two-phase flows are critical in various industrial settings, including nuclear reactors, heat transfer systems, chemical processes, and wastewater treatment. Air bubbles within these flows enhance mixing, enable oxygen transfer, and alter heat fluxes. In nuclear reactors, bubble dynamics and oxygen transfer play pivotal roles in containment cooling, pressure control, gas stripping, hydrogen/oxygen management, corrosion control, and thermal-hydraulic modeling, making a comprehensive understanding essential. Computational Fluid Dynamics (CFD) simulations offer powerful insights into these systems beyond what sensors alone can achieve.

This study examines the impact of two diffuser configurations on flow mixing and oxygen transfer in a 1.3-meter water-filled reactor with 16 air diffusers. Two configurations were tested: all diffusers active (configuration A) and only the central lines active (configuration B), both operating at a flow rate of 20 m³/h. Simulations using OpenFOAM's twoPhaseEulerFoam solver incorporated the two-film resistance model with Clift’s mass transfer coefficient. Experimental data collected on void fraction, bubble velocities, and liquid flow supported validation.

Findings showed strong alignment between simulated and experimental results for void fraction and velocity profiles, allowing for detailed analysis of flow patterns. Configuration B demonstrated a 15% reduction in oxygen transfer efficiency experimentally, while CFD predicted a 24% decrease, effectively capturing the trend. These CFD simulations offer pre-construction insights into diffuser performance, informing design decisions on hydrodynamic interactions and oxygen transfer efficiency. Future work will enhance model accuracy and explore additional flow rates and dynamic aeration configurations.

While many advanced reactor-design concepts do not rely on subcooled boiling, the Pressurized Water Reactor still dominates world-wide installed capacity and accurate prediction of transient and steady characteristics of the nucleate boiling heat transfer regime has a first-order impact on the reactor design efficiency and safety margins. After more than 70 years of study there remain gaps in knowledge and uncertainties in empirical models and correlations. With the continued increase in available computational power, interface resolving high-fidelity simulations have become an important tool in closing these gaps in knowledge. Numerical investigations at practical scales involving thousands of bubbles are now possible. However, resolving the micro scale surface topology and roughness necessary for in situ prediction of bubble inception and inertial growth remains computationally out of reach for the foreseeable future. In this work, we will present a closure model of vapor bubble nucleation waiting time and inertial phase growth aimed at reducing uncertainty in existing high-fidelity numerical investigations of nucleate boiling heat transfer. The model is based on a simplified energy and force balance on the extant vapor bubble retained in the micro-cavity. The inception of vapor bubble growth considers local thermodynamic effects and surface conditions and is formulated as a renewal time. Care has been taken to provide a numerically stable and computationally efficient closure to the higher-order thermal hydraulic simulations.

The motivation of the study of fully-implicit scheme for two-fluid model is, comparing with semi-implicit scheme, the mainly advantage of fully-implicit method is the restriction of the time step on stability is very small, and large time step is allowed. However, fully-implicit scheme makes the degree of coupling among equations is strong, and the solution become more difficult. First, the calculation of Jacobian matrix is difficult and highly time- and memory- consuming due to matrix is large. The Broyden-type method is adopted because it only requires calculate Jacobian matrix one time during the iteration process. We use numerical difference to estimate Jacobian matrix for the first iteration, then calculate Jacobian matrix by Broyden method for the remaining iterations. Due to the sparsity of Jacobian matrix, the Schubert method is applied. This method makes full use of Jacobian matrix’s sparsity. Second, the convergence performance become poor especially for strong interfacial effect case. The reason is the non-linear interfacial models makes the Jacobian matrix highly ill-conditioned. The relationship between condition number and the degree of interfacial effect is studied, and we try to reduce condition number by modify governing equation. Finally, this solver is tested. The calculation performance under very larger time step is assessed.

High-resolution simulations of two-phase flows can augment existing experimental data used for thermal-hydraulic system code development. PHASTA is one such high resolution code that uses Direct Numerical Simulation of the Navier-Stokes equations with the Level Set method to resolve individual bubbles and droplets of the flow. PHASTA, when deployed on high performance computing systems, has shown remarkable performance at simulating turbulent bubbly flows which occur in prototypical reactor geometries and conditions. However, flows of high void fraction systems present a challenge to the Level Set method which is well known for its mass loss deficiencies. High void fraction annular flows are typical near the top of boiling water reactor fuel channels and steam separators. Annular flows generate a large amount of small, entrained droplets and bubbles which require a prohibitively fine mesh in order to resolve and conserve the mass of these smaller objects. New numerical methods or models must therefore be incorporated into PHASTA in order to accurately model these flows with economical grid sizes. The Conservative Level Set method is one such method which has shown superior mass conservation properties on a variety of complex two-phase flows. This work describes the development and testing of the Conservative Level Set method implemented in the PHASTA finite-element code. A simulation of a swirling annular flow in a BWR steam separator is conducted to test the method on a large engineering-scale problem.

An important industrial issue and a continuing challenge in computational fluid dynamics (CFD) is the simulation of gas-liquid flows when several two-phase regimes coexist.

To overcome the computational challenges associated with all-scale interface resolving approaches, all-flow-regime CFD models have been proposed. One such method is the Generalized Large Interface Model (GLIM), implemented in the NEPTUNE_CFD software. This framework allows a smooth modelling approach transition: the small-dispersed scales are modelled, whereas the large scale gas-liquid interfaces are explicitly treated. These methods can be useful for many nuclear thermal-hydraulic applications.

To achieve this goal, GLIM and other models in the literature require specific closure terms for configurations where the gas-liquid interface is recognized. In addition to the interfacial closure formulations, blending functions and interface recognition methods are essential to deal with the transition between scales.

The aim of the present work is to evaluate different formulations of the interfacial forces. The NEPTUNE_CFD code is used. A rising large bubble configuration where the interface can be well resolved or rather diffused over several mesh cells was considered as a preliminary validation test. Several large interface drag formulations and cell number over bubble diameter ratios have been tested, the results are discussed in this paper. Then, the model was applied to simulate an intermittent air-water cross-flow in a tube array, featuring at the same time dispersed bubbles/droplets and large gas-liquid interfaces, a configuration close to the one encountered in U-tube steam generators.

Pool scrubbing is an effective process to decontaminate radioactive aerosols as severe accidents happen in nuclear power plants. Bubble residence time is one of the key parameters to determine the aerosol decontamination factor (DF) which is defined to describe the efficiency of aerosol removal, especially in the swarm flow region which makes a significant contribution to the total aerosol removal. The Euler-Euler two-fluid method and the Lagrangian Particle Tracking (LPT) method are applied, the former is used to get the flow field information, and the LPT method is used to track the bubble movement to obtain the bubble residence time. Through the analysis of bubble residence time distribution, the model of probability density function for bubble residence time is preliminarily established. The probability density function obeys well an exponential decay behavior. The decay constant is fitted according to the CFD simulation results. In general, the developed model shows good potential in predicting bubble residence time. Furthermore, the effect of bubble diameter on the probability density function is investigated. The bubble diameter shows a strong effect on bubble residence time, which is because larger diameter bubbles experience greater buoyancy, resulting in a higher rising velocity.

Simulating nuclear transients with Computational Fluid Dynamics (CFD) poses significant computational challenges due to complex physics and disparate temporal scales. These lead to high computational costs, necessitating advanced techniques for efficiency. This work explores two strategies: multi-rate timestepping with overset grids and reduced-order modeling (ROM).

First, we present an overlapping domain capability within NekRS, a GPU-accelerated spectral element CFD solver. This feature enables independent solution of spatial regions, improving efficiency for large-scale transients in complex geometries. We demonstrate its scalability through the TALL-3D experiment, a benchmark for thermal-hydraulic behavior in liquid metal reactors. Multi-rate timestepping significantly accelerates simulations, addressing a key CFD bottleneck.

Second, we develop ROMs as a computationally efficient alternative for high-fidelity transient simulations. ROMs support digital twin technologies, enabling rapid simulations with high accuracy. Using the NekROM framework, we implement Proper Orthogonal Decomposition (POD)-based ROMs to tackle challenges in nuclear modeling, including (1) thermal striping, which induces fluctuating thermal stresses affecting structural integrity, and (2) molten salt reactor (MSR) modeling, an emerging reactor technology. Both cases involve long time scales, benefiting from ROM acceleration. Our results show POD-based ROMs achieve significant computational speedup while maintaining essential accuracy.

These advancements in multi-rate time-stepping and ROMs represent a major step forward in transient simulation. By improving computational feasibility, they enable more efficient and accurate simulations of complex nuclear systems, enhancing reactor design, safety analysis, and operational decision-making.

In the realm of computational fluid dynamics (CFD), the high computational cost has always been a significant challenge, particularly in fields like nuclear safety analysis where complex flow problems are common. However, scientific machine learning (SciML) has demonstrated its effectiveness in solving real-world problems, including those in CFD. In recent studies, the issue of residual divergence in long-term simulations when using a single training approach has been identified. Our goal is to develop a flexible framework to optimize the state-of-the-art hybrid ML algorithm; residual based physics informed transfer learning (RePIT) developed by J. Jeon et al. This is a promising technique which has accelerated the simulation and ensures long-term stability using neural networks. However, its performance was only demonstrated on one network architecture known as finite volume method network (FVMN), also manual intervention was involved while switching between ML and CFD computation. Our projection is that we can further reduce the computational time by automating the switching process and also adding several network benchmarks where we could be able to implement state-of-the-art neural network architectures and choose the best performing one. In this work, we verified (1) the integrity of the framework by using the FVMN and compared the results with the original case study and (2) that various ML models could be loaded in the RePIT framework. In particular, DeepONet-RePIT shows the best acceleration performance. We believe that this hybrid approach is the most practical SciML utilization for robust CFD acceleration.

Typical prismatic core High Temperature Gas Reactor (HTGR) designs feature many individual coolant channels that are long in the axial direction and circular in cross section. When predicting fuel temperatures during initial concept design iterations, it is appropriate to model the steady, single-phase flow and heat transfer within each of these coolant channels using simple correlations. The challenge is then providing appropriate thermal boundary conditions along the walls of every individual channel in the core (e.g. accounting for power shapes). For complex core layouts, conjugate 3D CFD models of the entire assembly could be used to address this at the expense of significant computation time.

In this work we demonstrate a hybrid approach that uses a fluid-as-a-solid approximation to enable 3D simulations of entire core assemblies at reduced computational cost. This eliminates the need to solve the Navier-Stokes equations in the fluid, while leveraging existing functionality available within the CFD code STAR‑CCM+ to minimise implementation time and enable rapid design studies.

The approximation involves the fluid in each channel being modelled as having a suitable anisotropic effective thermal conductivity. This study considers gas flows within geometries relevant to prismatic HTGR cores. Predicted temperatures are compared between baseline CFD simulations and simulations employing the fluid-as-a-solid approximation. The fluid-as-a-solid approximation predicts similar temperature profiles to the baseline simulations. A reduction in computation time of around two orders of magnitude was achieved. This approach is expected to be valuable to preliminary concept designers, wherein rapid turnarounds are desirable across a range of designs.

In advanced reactor designs, dual-cooled annular fuel assembly has attracted significant attention due to its unique thermohydraulic characteristics. However, its complex structure poses challenges for the traditional analysis methods. In this paper, a Physics-Informed Coarse-Mesh method is proposed and applied to the analysis of annular fuel assembly. Given that annular fuel consists of internal and external channels, coarse meshes are employed to capture the primary geometric features, thereby limiting computational costs. Widely validated empirical correlations are used to correct wall friction and heat transfer, ensuring simulation accuracy. By developing a conjugate heat transfer method and a one-platform multi-scale coupling strategy with the fine-mesh CFD method, the issues of flow and heat distribution within annular fuel assembly are resolved. Based on the design parameters of the OPR-1000 reactor, a comparison of the thermohydraulic characteristics between cylindrical and annular fuel assemblies is conducted. The results show that annular fuel assembly exhibits lower central fuel temperature and higher pressure drop. The flow rate between different internal channels remains consistent. Furthermore, comparisons with existing programs demonstrate that this method can accurately simulate the flow and heat transfer characteristics of annular fuel assemblies, providing robust support for the design and optimization of advanced nuclear reactors.

The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) is a research reactor capable of delivering large-volume, high-flux thermal neutron irradiation in a realistic environment. Its design enables comprehensive studies on the effects of intense radiation on reactor materials and fuels. To ensure these experiments are conducted under specific conditions while maintaining safety standards, rigorous programmatic and safety analyses are required. These analyses typically consider coupled physics such as thermal-hydraulics, neutronics, structural analysis, and fuel performance. In this work, the Multiphysics Object Oriented Simulation Environment (MOOSE) framework and MOOSE-based applications are used for developing coupled multiphysics simulations for the ATR core. Regarding thermal-hydraulics analyses, traditional high-fidelity computational fluid dynamics (CFD) are often computationally expensive and the available codes do not have the physics models required for the simulation of the other aspects necessary for reactor analysis. This study leverages the coarse-mesh CFD capabilities in Pronghorn for conducting coupled thermal and neutronics analyses of the ATR core using a simplified porous-medium approach. This method homogenizes solid and fluid regions, enabling streamlined geometry and accelerated simulation times. This paper aims to: i) develop a 3D model of the ATR using publicly available data; ii) create a corresponding coarse-mesh CFD model; iii) verify simulation results against benchmark calculations; and iv) evaluate the use of the porous-media methodology for simulating the ATR. The results indicate that the coarse-mesh CFD capabilities provide accurate predictions for the temperature difference and pressure drop at the core, and fuel temperature distribution of the ATR, with improved run-time.

In the event of a severe nuclear power plant accident, radioactive materials may be released into the environment as aerosols and transported over long distances by atmospheric motion, posing significant risks to human health and the environment. In order to accurately characterize the temporal and spatial dynamics of radioactive aerosol dispersion after a nuclear power plant accident, a dispersion numerical simulation method based on the Euler-Euler model was proposed. An aerosol concentration distribution solver was independently developed based on the open-source computational fluid dynamics (CFD) platform to realize the numerical simulation calculation of aerosol concentration distribution in a large space. The stable release and dispersion process of cesium iodide (CsI) aerosol in different environmental wind fields was studied. The results showed that the aerosol spreads in the wind field near the ground, and the concentration was always high in the area within 400 meters from the source (the concentration was higher than 1.47 × 10¹⁶ m^-3). With 1.47 × 10¹⁶ m^-3 as the detection standard, in a 5m/s ambient wind field, aerosol spreads to 3660 meters downstream of the source at 800s. Assuming that the evacuation time is 10 minutes, the danger degree is highest within 1.77km of the source as the center, and residential areas are not suitable within 4km.

The SCENES program is an integrated software package for nuclear power plant design and safety analysis independently developed by Shanghai Jiao Tong University. In order to verify the accident analysis ability of its system analysis code SCENES-netFlow, this paper selects ACME bench for modeling analysis. The ACME test facility is based on CAP1400 and is mainly used to verify the safety of the passive system in the event of small break LOCA and non-LOCA in the prototype power plant. This paper mainly analyzes the working conditions of CAP03 small break LOCA. The results show that the predicted accident sequence and test phenomena are consistent with the experience. The main results of numerical analysis can well reflect the experimental phenomena and agree well with the experimental results, indicating that SCENES-netFlow has the ability to simulate the accident conditions.

In the event of a severe accident at a nuclear facility, molten core components, known as corium, can form and pose a risk if not properly cooled. Corium can breach the reactor pressure vessel and cause the ablation of containment concrete in a process called Molten Corium Concrete Interaction (MCCI). Understanding MCCI is essential for evaluating containment safety, with previous studies using experimental and numerical approaches. Traditionally, system codes and lumped parameter methods have been employed, while CFD simulations have largely focused on corium spreading. This study introduces a novel approach using a multiphase flow technique to predict natural convection and phase change processes in MCCI. The model integrates multiphase heat transfer, phase change, mass mixing, and decay heat generation, implemented in the OpenFOAM CFD code. It is validated against a PCM melting experiment, showing excellent agreement with experimental data. The validated model is then applied to simulate the COMET-L2 experiment, incorporating decay heat sources and phase changes. A mesh sensitivity study and time-step variations are conducted for model convergence, with results closely matching experimental data. Detailed analysis of concrete ablation, crust formation, oxide relocation, and metal penetration into the basemat is provided, offering insights into the thermal behavior of corium and concrete during MCCI. This approach enhances the understanding of MCCI phenomena and supports improved safety assessments for nuclear containment.

The Water-Cooled Lithium-Lead Test Blanket Module (WCLL-TBM) is an essential component in ITER that will provide crucial information for the development of the DEMO driver blanket. Our research aims to build a multi-scale system code for the thermal-hydraulic analysis of the WCLL-TBM. The OpenModelica software is used to develop a robust and modular object-oriented library for the components of the WCLL-TBM and the Water Cooling System (WCS) in this work. The various objects contain modelled thermal flow loops with 0D/1D interconnected components such as pipes, heat ports, orifices and valves. The objects describe the different multi-scales and can be nested and combined to form new objects. Such objects include the Double-Walled Tubes (DWTs), First Wall (FW), Breeding Units (BU), Breeding Module (BM), and larger outer circuits - all of which are designed to have replaceable modules with different levels of fidelity. The code aims at fast and reliable thermal-hydraulic predictions of the WCLL-TBM components and WCS during nominal operating conditions (gauge pressure 15.5 [MPa], inlet temperature 295 [textdegree C], outlet temperature 328 [textdegree C]), as well as the transient response of the system to off normal scenarios by varying certain parameters and loading conditions.

This research utilizes the system-level modeling language Modelica and its open-source libraries, Transform and Hybrid, to develop a natural circulation model for the primary loop of a Low-Temperature Nuclear Heating Reactor(NHR200-II). This model includes components such as the reactor core, coolant channels, heat exchangers , control system model and so on. Based on these models, the steady-state behavior of the primary loop under 100% nominal reactor power conditions was simulated. Also, transient simulation analyses were performed for step and ramp changes at 90% nominal power. The simulation results, when compared with RELAP5 data, demonstrated excellent agreement, confirming the validity and accuracy of using Modelica for simulation modeling. Furthermore, the primary control system model established in this study can regulate the core outlet temperature by controlling the reactivity of the core, and the results show that the reactivity control scheme is feasible. The research work in this paper lays a foundation for using Modelica language to carry out nuclear energy system simulation application.

Different types of medical isotopes are needed for kinds of procedures where many options for production are possible. One particular isotope, Astatine-211 or At-211, can be produced using cyclotron based irradiation where a Bismuth target is converted to At-211. During irradiation, significant heat is generated within the target and appropriate cooling is needed to prevent target melting and increase isotope yield. In support of higher At-211, University of Washington (UW), Oak Ridge National Laboratory (ORNL) and Virginia Commonwealth University (VCU) are collaboratively developing better At-211 target and target holder designs to enable higher isotope production yields. In this study, VCU is focused on the thermal-hydraulics modeling of different target designs including material and geometric parameters to enable UW and ORNL collaborators to hit desired yields. Initially VCU is focused on 0-D modeling using lumped parameter analysis approaches to enable a design space to be developed using multi-objective optimization. This enables the ability to explore both differential holder materials (e.g. aluminum or stainless steel varieties) and coolant channel geometries rapidly to reduce the total number of high-fidelity CFD simulations and experiments. The 0-D model was created in Python to include two energy balance ordinary differential equations (ODEs) to predict the Bismuth and sample holder temperatures during irradiation. The heat generation within the Bismuth target and the coolant conditions were acquired from the UW team and implemented in the model. Using the 0-D model,, we’ve identified several potential improvements to both the sample holder design and coolant channels for follow-on CFD and experimental studies.

Fluid-structure interactions play a critical role in numerous engineering applications, such as jet flows around fuel rods in nuclear reactors. Under specific flow conditions, these interactions can give rise to vortex-induced vibrations (VIV), a phenomenon where large-amplitude oscillations occur due to vortex shedding. VIV poses a significant threat to system stability and can lead to operational failure. Therefore, understanding and controlling VIV is essential to mitigate its detrimental effects.

This study explores the passive control of VIV in a circular cylinder that oscillates freely, using a non-linear energy sink (NES). The NES is designed as a secondary system incorporating linear damping and a key non-linear cubic stiffness component. Simulations are conducted using the Reynolds-averaged Navier–Stokes (RANS) turbulence model with strongly coupled fluid-structure interaction model, utilizing the dynamic response of both the cylinder and the NES, as well as the surrounding fluid flow. By systematically adjusting the sink parameters—mass, spring, and damping—this research investigates their influence on the behavior of the coupled system. The system's response is analyzed at reduced velocity within the lock-in range, where the cylinder's motion synchronizes with vortex shedding. The model is validated against existing data from literature which indicate the optimum values for these parameters to achieve the best performance. Key results are presented in terms of vibration amplitude, drag and lift coefficients, Strouhal number analysis, and vortex visualization, providing insight into the effectiveness of the NES in controlling VIV.

Solid-solid contact thermal conduction is a basic heat transfer problem in thermal power engineering, which is significant for thermal design and safe operation of system equipment, such as high temperature thermal protection of aircrafts, efficient thermal management of space orbits, and superior heat transfer chain of nuclear engineering. Objective to the solid-solid contact thermal conductivity of typical structure based on tubes and holes in heat-pipe nuclear reactor systems, theoretical models of thermal conductivity, mechanics and thermodynamic coupling at the microscopic contact interface were established in the paper, obtaining the interface contact thermal conductivity characteristics under the filling of helium in micro gaps, as well as the influence on the contact thermal resistance for interface temperature and external loads. According to the CFD simulation results under different interface temperature and external loads, the Levenberg-Marquardt algorithm was used to fit a high temperature contact thermal resistance correlation on the cylindrical interface. By selecting appropriate fitting parameters, the R-squared corresponding to the fitting results was greater than 0.95, indicating that the calculation model had a good predictive ability for the contact thermal resistance. It was applicable for the rapid evaluation of contact thermal resistances for the solid-solid interface in engineering design and heat transfer analysis of heat pipe reactors.

The fuel irradiation device serves as a critical platform for conducting nuclear fuel irradiation experiments in research reactors. Its structural design, thermal characteristics, and the arrangement of measurement points at the outlet significantly influence experimental results, thereby affecting the thermal power determination of the device and the evaluation of fuel performance. This study focuses on the HFETR irradiation device, employing a CFD-based three-dimensional high-resolution modeling method to investigate the impacts of outlet sensor placement and structural thermo-mechanical properties on power measurement accuracy. Computational results demonstrate that positioning temperature sensors 130 mm upstream of the physical outlet plane effectively represents the outlet temperature field. From a thermal-hydraulic perspective, an annular gap thickness of 1 mm achieves a coolant flow partitioning of 22% through the bypass channel, with parasitic heat losses limited to 4.7% of the total generated power. This configuration ensures adequate cooling of the samples while avoiding excessive heat leakage.

Helical cruciform fuel (HCF), a novel nuclear fuel design, shows potential for enhancing power output and extending the service life of light water reactors (LWRs). While thermal-hydraulic studies on HCF assemblies are common, coupled analyses with high-fidelity neutron physics remain limited. This study establishes a CFD model of a 3×3 HCF assembly, integrating volumetric heat sources derived from neutron physics calculations to investigate flow and heat transfer phenomena.Key findings include helical variations in heat flux density (q) and wall temperature (T_w) along the flow direction. Due to the gap effect, q is lower in valley regions compared to blade regions, while T_w shows the opposite trend. Bulk temperature (T_l), however, lacks noticeable helical patterns. Under different heat source conditions with identical total power, peak values and positions of q, T_w, and T_l vary significantly. Condition one results in 70% higher q, an 18.8 K rise in T_w, and an 8 K increase in T_l compared to condition Two, with peaks occurring in different axial regions. Conversely, condition Two shows minimal axial q variation, with T_w and T_l peaking at the outlet. These results suggest a higher likelihood of boiling crises under condition one. Increased local axial power exacerbates circumferential non-uniformity in q and T_w, with heat transfer deteriorating at blade regions aligned with Twist angles of multiples of 90°, marked by reduced q and elevated T_w and T_l.

Fuel elements play a vital role in the safety and economic efficiency of nuclear power plants as one of the core components of a reactor. In addition to facilitating inter-channnel mixing between rods and enhances heat transfer, the wire-wrapped rod bundle is free from the supporting of spacer grid due to its self-locating structure through the contact between adjacent rods, which making it popular in recent research on reactor structural design. However, complex mechanical interactions often occur in wire-wrapped fuel rods under the constraints of reactor irradiation and high temperature, leading to stress concentration at the contact points of adjacent rods. This can easily cause fatigue damage to the fuel clad and affect its integrity. This study establishes a finite element model of wire-wrapped fuel rods by using the Ansys Workbench, taking into account the effects of irradiation, high temperature, and the geometric structure of the wire. The mechanical interaction characteristics of wire-wrapped rods under complex working conditions are investigated. The results obtained from this study on the mechanical characteristics of wire-wrapped rods can provide insights for the structural optimization design of fuel rods.

Externally cooled dead-end pipes, thermally and hydraulically connected to a hot source, exhibit complex physics, primarily driven by the interaction between penetrating swirl at the open end and stratified flow near the closed end of the pipe. These inherent instabilities in practice can lead to temperature fluctuations, potentially causing thermal fatigue and leakage in stainless-steel pipes. Such uninsulated pipes may be found, for example, in some 2 loop Westinghouse PVRs, where the safety injection (SI) pipes are connected directly to Reactor pressure Vessel (RPV). To better understand the thermal-hydraulic behaviour of this SI pipe configuration, CFD simulations were conducted. Despite advancements in computational power, such industry-level simulations remain challenging due to numerous uncertainties affecting prediction accuracy. Our study highlights that prediction of the turbulent swirl plus competing with the natural circulation is highly sensitive to CFD model settings, including boundary conditions, mesh resolution, turbulence models, and numerical methods (e.g., discretization schemes, solver types). These sensitivities suggest that the phenomena are unstable and chaotic. External air cooling, which induces flow stratification, emerges as the primary source of uncertainty, while unknown geometry at the inner edge of the DVI nozzle where the swirl forms, adds further complexity. Additionally, the large geometrical model restricts a systematic mesh sensitivity study, and the lack of sufficient experimental data limits the validation of turbulence modelling. All the above-mentioned aspects will be systematically reviewed in our paper, and supported by the CFD examples.

Integral effect test data obtained from the High Temperature Test Facility (HTTF) are being used for benchmarking CFD and system codes in the OECD-NEA Thermal Hydraulics Code Validation Benchmark for High-Temperature Gas-Cooled Reactors using HTTF Data. Five system codes SAM, RELAP5-3D, GAMMA+, SPECTRA, ARIANT, and CATHARE are used to model benchmark Problem 3 Exercises 1C and 1D, which simulate the steady state and pressurized conduction cooldown (PCC) transient of HTTF Test PG-27. This test examines the PCC phenomena progression in an integral test facility scaled to the General Atomics MHTGR design. The proposed exercises include well defined boundary conditions and assumptions so that code-to-code comparisons will help identify differences between modeling approaches, numerical methods, and uncertainties in the solutions of different codes. Preliminary assessment of the results shows that in steady-state operating condition, the codes agree very well for key parameters such as coolant temperature, solid temperature and flow distribution in core regions. In PCC transient, the agreement is reasonably good. The codes predict that natural circulation is several orders of magnitude lower than steady-state flow rate and core-wise heat transfer is therefore dominated by thermal conduction and radiation. The codes also predict similar temperature trends in the solid structures but there are discrepancies in the transient behavior. This is not surprising considering the vastly different modeling schemes of a very complex core geometry.

Extreme accidental scenarios in nuclear power plants (NPPs) may involve hydrogen formation and its pressurized release into the containment building, potentially leading to unintended explosions. A fundamental understanding of the complex physical mechanisms associated with such scenarios is critical for their prevention and mitigation. This includes investigating hydrogen jet dynamics during the initial stages of release under high-pressure conditions, which are relevant for hydrogen storage systems in nuclear facilities.

This work uses high-fidelity 3-D numerical simulations based on the Implicit Large Eddy Simulation (ILES) technique to investigate the turbulent characteristics and mixing of underexpanded jets, varying initial pressure ratios, and jet diameters. First, nitrogen jets released into atmospheric nitrogen are investigated, examining pressure ratios of 60, 30, 15, and 7.5 for a 3 mm diameter jet. This serves as a simpler case to analyze the jet flow dynamics. Second, we focus on high-pressure hydrogen jets released into air with the same pressure ratios and different jet diameters of 1.5, 3, and 6 mm to represent a reactor-scale problem. Both cases are validated against experimental data with an excellent agreement.

Key findings include the influence of pressure ratio and jet diameter on the turbulent jet self-similarity and mixing shear layer dynamics. A larger jet diameter enhances self-similarity, while a decrease in pressure ratio disrupts it. Higher pressure ratios result in thicker shear layers and broader temperature ranges. These insights contribute to enhancing safety procedures and protocols in nuclear systems and other high-pressure hydrogen storage applications.