Conference Agenda

^{The safety analysis of nuclear reactor containment is crucial for maintaining the integrity of nuclear power plants during accident scenarios that threaten containment integrity. The Fukushima-Daiichi incident underscored the importance of containment as the ultimate barrier against the release of radioactive materials into the environment. During a loss of coolant accident (LOCA), the release of coolant from the reactor coolant system (RCS) elevates the temperature and pressure within the containment. Investigating these parameters is vital for ensuring the containment wall’s integrity. In this study, an Unsteady Reynolds-Averaged Navier-Stokes (URANS) simulation was conducted to examine the steam-air mixing behavior inside the containment during a LOCA. The steam injection nozzle is located in the lower part of the steam generator SG-2 compartment. The instantaneous temperature profiles of the steam-air mixture, predicted by various turbulence models, were validated against experimental data at different locations within the containment. The numerical predictions showed good agreement with the experimental temperature profiles. Additionally, the impact of LOCA steam injection on the compartment and containment walls was investigated. The numerical investigation revealed a significant impact of steam injection in the SG-2 compartment and the lower section of the containment. Furthermore, steam was found to be uniformly stratified in the upper section of the containment, exhibiting a comparatively lesser impact from the steam injection during the early transient phase."}

The reflood phase of a loss-of-coolant accident in a nuclear power plant is crucial for safety, as it determines the peak cladding temperature. A high accuracy in the prediction of this parameter by thermal-hydraulic system codes like RELAP5 and TRACE is essential to ensure compliance with regulatory safety limits. Experimental programs, such as the Rod Bundle Heat Transfer (RBHT) project, provide benchmark data for evaluating and improving these models.
This paper assesses the ability of TRACE to simulate the reflood phase, focusing on postcritical heat flux phenomena in the upper rod bundle regions. A TRACE model was developed, incorporating key features like the spacer grid model and a droplet field. The results were compared with experimental data and previous RELAP5 simulations. Additionally, a sensitivity analysis examined which parameters have a stronger influence on the PCT and the quench time.
The findings indicate that TRACE performs well in high reflood rate scenarios but struggles with low or variable flooding rates, particularly in capturing heat transfer mode transitions. The sensitivity analysis highlighted key factors affecting simulation accuracy, such as the quench temperature and the heat transfer coefficient during dispersed flow film boiling. Despite some limitations, TRACE demonstrated potential in modeling certain phenomena, including heat transfer enhancement from spacer grids. Further improvements could enhance its predictive capabilities, increasing confidence in its use for safety assessments of nuclear reactor reflood scenarios.

Sodium-cooled fast-neutron reactors are currently considered to be the most mature type of reactor able to closing the fuel cycle. In France and throughout the world, pool-type reactors are selected to build generation IV power plants. In a sodium-cooled pool-type reactor, thermal stratification can occur in the pools in several cases. This phenomenon is monitored closely because it can impact the behaviour of the reactor and might lead to thermal fatigue. In the 1980s, the SUPERCAVNA test facility was operated at the CEA Grenoble research centre. The experimental campaigns investigated the onset of thermal stratification in a rectangular pool. During transient tests, cold sodium was injected in a hot sodium pool. Depending on the inlet flow velocity, thermal stratification would form and erode the hot sodium layer in the pool. The data from these tests constitute a set of CFD-grade experiment that are very useful to assess the capability of CFD codes to capture the relevant phenomena. Code_Saturne was selected to perform calculations of three transient tests from the SUPERCAVNA experimental campaign. Two tests were well captured. A mixed convection test proved more difficult to predict and lead to extensive tests of turbulence and turbulent heat flux models. In this paper the SUPERCAVNA facility and the tests of interest are presented. Then, the CFD model of the facility is described and the results are presented and discussed. Conclusions and recommendations for this type of flows are proposed.

Nuclear thermal hydraulics codes, such as TRACE, frequently exhibit a lack of thorough documentation regarding their input parameters, particularly for crucial elements like heat transfer coefficients, which are often determined through expert judgment and empirical correlations integrated within the code itself. To enhance the reliability and precision of simulations conducted with these codes, it is vital to systematically assess the uncertainties tied to these parameters. TRACE (version 5.0 Patch 8) supports this by enabling users to adjust 43 physical model parameters in the input script using multipliers, starting with a default value of 1. In this investigation, we conducted a tube condensation test at KAIST's Passive Containment Cooling System Facility using TRACE, followed by a detailed Sensitivity Analysis (SA) aimed at identifying and ranking the parameters that significantly affect a key output variable—the saturated steam temperature at the center—critical for Inverse Uncertainty Quantification (IUQ). We developed a robust mathematical framework for Maximum Likelihood Estimation (MLE) based on the Expectation Maximization algorithm and applied it to the tube condensation test. The sensitivity analysis identified several parameters that influence the code’s predictions of the saturated steam temperature, with the vapor-to-interface and liquid-to-wall heat transfer coefficients being the most impactful. Following this, we implemented inverse uncertainty quantification to assess statistical properties like mean and variance for these key parameters. The MLE approach provided reliable estimates of their probability density functions, significantly enhancing our understanding of the uncertainties involved in TRACE simulations.

In the frame of the European McSAFER project, experimental and numerical investigations for the safety evaluation of water-cooled SMRs such as NuScale, SMART, CAREM and F-SMR were performed. At LUT, the MOTEL facility was designed based on the NuScale geometry with buoyancy driven primary circuit flow including. Experiments with asymmetric heated core for crossflow studies were challenging for numerical simulations because of flow instability and correct prediction of mass flow and pressure loss. At KIT, detailed CFD models for the entire vessel with all components of the primary circuit were developed. The best suited CFD model version was resolving all primary loop components like the core heater rods with spacer grids and the helical coiled heat exchanger tubes of the steam generator in detail. Therefore, more than 2*10⁸ cells were used. A detailed analysis of the simulations and experimental data demonstrated the necessity that also even parts of the SG secondary circuit containing two-phase flow has taken into account in order to obtain full agreement with temperature measurements. Furthermore, several CFD models with simplifications such as modelling the SG´s by a porous media or the consideration of the full resolved core region as a standalone part with specified inlet and outlet conditions were created. The deviations between experimental data and the various model simulations clearly demonstrates the disadvantages of model simplifications and justifies the numerical costs of a detailed full vessel CFD model, which provided very good predictions.

In sodium cooled fast reactors (SFRs), transient thermal stratification of the sodium coolant within reactor components must be thoroughly understood for safety analysis and licensing efforts. Computational fluid dynamics (CFD) modeling can be leveraged to simulate the thermal hydraulics within these reactors and study the transient thermal stratification of their coolant media. However, before they can be used for safety analysis and licensure, these models must be experimentally validated to ensure their results are consistent with physical observation. The Gallium Thermal-hydraulic Experiment (GaTE) studied thermal stratification within SFR upper plena, utilizing liquid gallium as a surrogate fluid for the liquid sodium. Cold-shock flow injection tests conducted with GaTE provide an experimental benchmark for validation of CFD in capturing thermal stratification within SFRs. Large-eddy simulation (LES) of the cold-shock tests of GaTE facility was conducted with Nek5000. The velocity response along the plenum height during the isothermal stage prior to the shock is validated against the GaTE experimental benchmark. Then two mixed convection regimes were simulated, one with more dominant effects of forced convection and one with more dominant effects of natural convection. The axial temperature profiles within the plenum during the thermal transient are then compared to those collected with GaTE. When validated, these LES models can be used to augment and extend the current understanding of thermal stratification within SFR plena by exploring a broad range of convection regimes. These validated data can be used to develop reduced-order models and investigate underlying turbulent mechanisms of transient thermal stratification.

International Atomic Energy Agency conducts Coordinated Research Project (CRP) on “Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop (NACIE)”. The project is a benchmark simulation of three experiments from NACIE lead-bismuth eutectic (LBE) experimental loop located at the ENEA Brasimone Research Center, in Italy, which includes simulation of transition from forced circulation by gas lift-off pumps to natural circulation. The experimental data from the tests, along with other benchmark specifications, was provided to the CRP participants for analysis with computational codes. The project work is organized in the five work packages (WP), of which WP1 is the system thermal hydraulics. In this work package, participants do calculations and submit the result of the test steady-state and transient simulations with system level codes. The codes used by the CRP participants for the WP1 simulations include: FRTAC, LOCUST, THACS, TRACE, RELAP5, RELAP5-3D, GAMMA+, SPECTRA, THOR, HYDRA-IBRAE/LM, and SAS4A/SASSYS-1.

This paper presents the collective results from all WP1 participants for the benchmark tests, as well as comparison with the experimental data. The results of interest for this package include the void fraction in the gas lift-off region from gas bubble, pressures and temperatures along the loop, and the LBE flow rate. The comparison is presented for the steady-state result at the beginning and end of the test, as well as for the transient results. Several conclusions are drawn from the collective comparison, mostly in terms of where the particular models are different from other codes or the test data.

The Rod Bundle Heat Transfer (RBHT) facility provides advanced and detailed experimental data on coolant flow and heat transfer in a 7x7 fuel bundle model under reflood conditions. Tractebel participated in the first phase of the project (RBHT-I) from 2019 to 2022, using the subchannel code CTF to model reflood conditions at low pressure, with varying flow rates, average power, and subcooled core inlet temperatures. This investigation revealed deficiencies in the CTF reflood model, highlighting the need for improvements, particularly in the flow regime map and the entrainment model. RBHT experiments offer a valuable database for code validation and are utilized in the current investigation to assess enhancements in the latest version of CTF’s reflood models. System codes such as CESAR are also evaluated. The newly implemented models in CTF show improved agreement with experimental data in some cases, especially regarding quenching time. However, they still tend to overestimate the Peak Cladding Temperature and predict a delayed quenching front. CESAR calculations, when coupled with a thermo-mechanics module in the DRACCAR Multiphysics platform (MP), demonstrate high sensitivity to initial conditions, such as the initial rod temperature.

The advanced system analysis tool, System Analysis Module (SAM), and subchannel code MOOSE-Subchannel are both developed under the U. S. DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. To improve the prediction accuracy in the reactor fuel assemblies, SAM and MOOSE-Subchannel are coupled with domain overlapping coupling approach. In the coupled simulation, SAM system model provides the overall system behavior, while MOOSE-Subchannel model provides more detailed solution within an assembly. Natural Circulation Experiment (NACIE) facility was a lead-bismuth eutectic (LBE) experimental loop located at the ENEA Brasimone Research Center in Italy, for study of the thermal-hydraulics behavior of LBE in rod bundle configurations. The NACIE loop includes a fuel pin simulator (FPS), which consists of 19 wire-wrapped electrically heated pins. The instrumentations of NACIE can provide temperature at different locations, mass flow rate, pressure during transient tests. In this study, a coupled SAM and MOOSE-Subchannel model of the NACIE loop is developed and benchmarked against experimental measurements. The temperatures from the coupled calculation agree well with the experimental data from thermocouples at various locations, including different points of the loop and local subchannel positions in FPS. Furthermore, the coupled SAM and MOOSE-Subchannel simulation results are compared with SAM standalone results, in which the FPS region is modeled using single-channel or multi-channel approaches.

Core reflooding, the injection of water into the reactor core during a Loss-of-Coolant Accident (LOCA), is a critical Accident Management strategy for water-cooled reactors. As part of the PERFROI project, the COAL experiments were designed by IRSN (now ASNR since January 2025) to study the coolability of intact and partially deformed fuel assemblies under reflooding conditions. Following this program, OECD/NEA/CSNI launched the International Standard Problem (ISP-53) based on the IRSN reflooding COAL experiments, which started in February 2024. This benchmark aims to evaluate the predictive capabilities of computational codes against COAL experimental data, focusing on undeformed and deformed fuel rods. ENEA is contributing to the international benchmark with the two codes DRACCAR and ASTEC. While ASTEC employs a simplified 2D core geometry designed to simulate a comprehensive Severe Accident scenario, DRACCAR features a more detailed 3D core representation with detailed thermo-mechanical modelling of fuel rods. Both codes share CESAR thermal-hydraulic module and include comparable models for reflooding and thermal-hydraulic related phenomena. This paper presents simulation results from both codes for two COAL experiments involving reflooding of a 7x7 undeformed fuel bundle. A detailed comparison against experimental data, along with key thermal-hydraulic parameters analysis, highlights performances and predictive capabilities of the two codes.

When a break occurs in a nuclear reactor, a fast cooldown has to be down to prevent the melting of the core. This is done by the injection of cold water at 7°C, in a pressurized vessel at 295°C. This is a Pressurized Thermal Shock. To improve the safety of the nuclear reactor, the EDF R&D team built an experimental facility in order to analyse the mix of hot and cold water in the downcomer of a 1300 MWe French Pressurized-Water-Reactor. This is the HYBISCUS-II experiment. Salt water at 45°C was injected into stagnant pure water at around 15°C to represent the injection of cold water in hot water. A scaling has been established to link the experiment to the reactor case. Here, we present a numical simulation of the HYBISCUS-II facility, made within the CATHARE code. We use a second scaling, developped for the BORA4-PTS experiment, in order to compare the numerical and the experimental results. The numerical simulations gives results that shows less than 1°C of difference with the experimental ones. With this experiment, we demonstrate the excellent capacity of CATHARE for the modelisation and simulation of a downcomer in a Pressurized Thermal Shock situation.

Within the frame of the European project PASTELS, which aimed to improve understanding of passive safety systems for PWR applications, several experiments were carried out and analyzed numerically. These experiments studied passive systems such as the Safety Condenser (SACO) or the Containment Wall Condenser (CWC). The new databases acquired during the project were simulated by the various project partners using simulation tools at system scale, lumped parameter codes for severe accidents or CFD.

The article will focus on the modeling of the PASI experiments representing a CWC in an open loop configuration. The experimental facility consists of a thermosyphon loop connected to a water pool at ambient pressure located in the upper part, and heated in the lower part through a tubular heat exchanger placed in a vessel which acts as the reactor containment.

A series of ten tests was interpreted by seven different organizations using four system codes (CATHARE, ATHLET, RELAP, TRACE) and two severe accident codes (MELCOR, ASTEC).

The article will briefly present the experimental set-up and the tests carried out. The modeling challenges will then be detailed: the physical phenomena such as condensation, flashing, dynamic instability, thermal stratification, and the different geometric domains to be represented (pipe, heat exchanger, volumes). Finally, the article will present different strategies to model these interrelated phenomena, and will discuss the main results, with a particular focus on two phase flow instabilities.

The forensic analysis of Fukushima Daiichi have underscored the significant role played by Units 1-3 suppression pools in attenuating Source Term to the environment. However, such a role is not limited to Boiling Water Reactors (BWR), but extend to accident scenarios in Pressurized Water Reactors (PWR), like Steam Generator Tube Ruptures (SGTR). The fission products trapping in water ponds, also referred to as pool scrubbing, was profusely investigated in the 80’s of last century and, eventually, encapsulated in stand-alone codes, like SPARC90, which formulation was embedded later in MELCOR. Despite the investigation effort, which was revived by the Japanese accidents, the diversity of accident scenarios involving pool scrubbing and the extraordinary complexity of its multi-physics nature, have made it hard to build a complete and sound database. As a consequence, modelling are still open for further development, testing alternate approaches and, awareness of uncertainties affecting pool scrubbing estimates.

This keynote lecture reviews the status of knowledge, pinpoints gaps necessary to be addressed and proposes a path to do it.

Due to its high latent heat, pool boiling exhibits excellent heat dissipation capability and is widely used in many industries such as electronic devices, steam generators, nuclear reactors, etc. This paper experimentally investigates the heat transfer enhancement effect of bi-conductive surfaces in pool boiling and reveals its underlying mechanism.Because of the enormous difference in conductivity, the heat transferred through the low-conductive epoxy can be ignored, and boiling only occurs on the high-conductive copper surface. This provides a method that can create specifically appointed spatial surface temperature variations and induce ordered liquid and vapor paths by changing the pattern of epoxy.This paper designs two types of epoxy patterns with a fixed copper area ratio of 60%, including reticular epoxy samples and more complex reticular epoxy with squares in the middle samples. At the same time, the number and width of stripes in the reticular epoxy are changed in each category. According to the result, almost all reticular epoxy samples enhance the CHF compared with bare copper. With the increase in the number of stripes in the reticular epoxy, the CHF displays the tendency to rise at first and decline in the end, reaching the peak of 77.2% CHF increase. Under the same number of stripes, reticular epoxy with squares in the middle samples have higher HTC and CHF compared with reticular epoxy samples because of superior wetting effect of epoxy. The results also suggest that overly narrow epoxy can lead to insufficient wetting effect and trigger CHF in advance.

During flow boiling, two different regimes are observed namely nucleate boiling and convective flow boiling. Nucleate boiling is dominant at high heat fluxes where bubbles produced at the wall are attributed the control of the heat transfer. Convective flow boiling is dominant at low heat fluxes and the heat transfer coefficient is observed to be directly dependent on the mass flux and the thermodynamic quality. The transition between to these two regimes has motivated vast research to determine if the transition is triggered sharply or there is a region where both mechanisms are presented. In this work, we study the nucleate flow boiling to convective flow boiling transition experimentally in a horizontal heated pipe.

Accurate prediction of heat flux is essential for various industrial and scientific applications, particularly in heat transfer and thermal management systems. Traditional methods for heat flux estimation often rely on complex physical modeling and intrusive sensor-based measurements, limiting their applicability in dynamic boiling conditions. Recent advancements in deep learning, particularly Convolutional Neural Networks (CNNs), have enabled non-intrusive heat flux prediction directly from boiling images. In this paper, we propose a multi-branch convolutional neural network architecture for heat flux prediction from boiling images. The key innovation lies in the introduction of multi-branch convolutional components (MB-Conv), which integrate multiple convolutional branches with varying kernel sizes to extract a comprehensive set of features.

The proposed model leverages a multi-branch architecture to enhance its representational power, allowing it to effectively learn from a diverse range of spatial features that are critical in predicting heat flux in boiling systems. Experimental results demonstrate that our model significantly improves prediction accuracy compared to the traditional single-path convolutional neural model. Moreover, the proposed multi-branch architecture outperforms several well-established CNN models, such as AlexNet, VGG, ResNet, DenseNet and EfficientNet in terms of predictive performance, highlighting the effectiveness of our approach in capturing the complex thermal phenomena present in boiling images. The ability to predict heat flux from boiling images opens up new possibilities for optimizing system performance and ensuring safety in thermal systems.

Enhancing the heat transfer is of interest for a wide spectrum of industries to achieve higher thermal efficiency. During subcooled flow boiling, liquid-to-vapour phase change causes high heat transfer rates, although the coolant bulk temperature is below its saturation temperature. When the required wall superheat is developed, vapour bubbles form over the heated surface marking the onset of nucleate boiling. When a bubble sufficiently grows in size, it departs from the heated surface, and its departure size and frequency dictate the enhancement of heat transfer rates. Formation of larger bubbles near the heated surface may result in their coalescence forming a local dry patch which may eventually lead to Critical Heat Flux (CHF). Although there are numerous correlations available in the literature, to estimate the bubble departure size, most of the correlations perform well at high pressure conditions. To this end, it is important to study the bubble departure size and its influence on the subcooled flow boiling characteristics at low pressure conditions. In the present study, Eulerian-Eulerian multiphase flow (EEMF) model is employed to simulate the subcooled flow boiling conditions. Instead of using an existing empirical correlation, a force balance model is developed to predict the departure size and validated against the experimental data at low pressure conditions. Based on this model, a correlation for departure diameter is developed which is coupled with the EEMF model framework. The developed correlation is found to be more accurately capturing the local vapour volume-fraction profiles compared to the existing correlations.

We employ three fast and synchronized optical techniques (white-light interferometry, infra-red thermography, shadowgraphy) to study the near-wall phenomena during the growth of a single bubble in saturated pool boiling of water at atmospheric pressure. Our focus is on the impact of applied heating on bubble growth dynamics, as well as the near-wall features: dry spot spreading, the liquid thin film (microlayer) that can form between the heater and the liquidvapor interface of the bubble and the interfacial thermal resistance. We found that varying the applied heating power does not significantly impact the bubble macroscopic and near-wall features. It is explained by large heat capacity of the heater. The only affected parameter is the waiting time, which decreases with the applied heating power. The interfacial thermal resistance shows no dependence with heat flux, and increases monotonously over time due to the progressive accumulation of impurities at the interface. We show that the triple contact line dynamics depends on the wall superheating at the contact line.

Microlayers beneath nucleating vapour bubbles are pivotal in enhancing bubble growth through evaporation during boiling, making their formation and depletion critical for accurate boiling heat transfer predictions. Recent studies employing advanced techniques such as Synchrotron X-ray imaging and Direct Numerical Simulations (DNS) have revealed significant morphological variations in microlayers during nucleate pool boiling on micro-structured surfaces. Bubble coalescence, a common phenomenon in nucleate pool boiling, further complicates microlayer dynamics. This study addresses a commonly observed but poorly understood coalescence event, where an ejecting bubble merges with a nucleating bubble on a micro-structured surface. Leveraging Synchrotron X-ray imaging of nucleate pool boiling and DNS of the bubble merging process, we report a jet formation mechanism induced by such a coalescence phenomenon, which leads to the rapid depletion of the microlayer. These findings provide essential insights for improving the accuracy of boiling heat transfer predictions.

Corrosion of structural materials is a critical issue for the nuclear industry. A major challenge concerns devices operating under boiling conditions, where corrosion can be influenced by various factors, in particular wall temperature, heat flux at the wall, and boiling. To enable an accurate modeling of corrosion in such industrial conditions, a comprehensive understanding of bubble behavior and associated thermal characteristics is imperative. This study aims to investigate the bubble dynamics and heat transfer through direct numerical simulations using the well-validated open-source code TRUST/TrioCFD. In this study, two-dimensional axisymmetric simulations are performed to investigate the growth and departure of bubbles originating from a single nucleation seed, focusing particularly on the effect of micro-region adjacent to the liquid-vapor-solid triple contact line and the transient conjugate heat transfer between the fluid and the adjacent solid wall. A multiscale modeling approach is adopted. The CFD-algorithm at the bubble-size scale is coupled to a sub-grid micro-region model. The micro-region model describes the partial wetting case. It takes the wall superheating and microscopic contact angle as inputs, and predicts the apparent contact angle and heat flux. Entire boiling cycles, including growth and departure phases followed by a waiting period, were simulated. We obtained detailed information on wall surface temperature and heat flux, directly applicable in corrosion models. Sensitivity analysis to the wall properties demonstrated that materials with higher thermal diffusivity exhibit a larger apparent contact angle, longer growth time, larger departure diameter, and shorter waiting time.

Dynamical System Scaling (DSS) is an innovative scaling methodology focused on incorporating transient behavior in the scaling criteria. This paper applies DSS to bubble dynamics by comparing experimental data to analytical solutions of the Rayleigh-Plesset equation. The Rayleigh-Plesset equation governs the motion of bubbles in an infinite body of fluid, and it is derived by simplifying the Navier-Stokes momentum equation with spherical symmetry. The motivation of this study is to evaluate a bubble growth correlation for rapid vaporization transients, which was identified in a previous DSS analysis with a peculiar initial growth rate. A simplified Rayleigh-Plesset equation is derived assuming that the bubble growth is solely inertially controlled. Atomic bomb test data is used for comparison purposes as it is presumed to have negligible heat transfer on the shockwave interface. DSS is employed to calculate the distortion between the experimental data and the analytical solution. Additionally, results will be compared to theoretical DSS work previously performed which applied DSS to bubble dynamics. This asymptotic analysis concludes that the correlation is unsupported, and the original test data does not cover the early stage. Therefore, DSS successfully identified that further improvement is needed to the existing correlation.

Small Modular Reactors (SMRs) constitute a significant advancement in nuclear technology, where Pressurized Water Reactors (PWRs) are extensively applied. Within PWRs, pressurizers play a critical role in maintaining pressure and help ensuring thermohydraulic safety. Nitrogen gas pressurizers offer advantages such as rapid response, compact design, and simplicity, rendering them more suitable for SMRs than steam pressurizers. Nonetheless, nitrogen may dissolve in cooling water and desorb as bubbles during pressure transients, leading to two-phase flow and damage reactor safety. Currently, a substantial gap exists in the accurate prediction of the conditions under where and when bubbles nucleate, how they evolute and impact on thermohydraulic safety. To establish a reliable predictive model and identify solutions for enhancing reactor safety, it is imperative to investigate the nucleation and growth behaviors of bubbles under supersaturated conditions during depressurization. This study aims to elucidate the supersaturation ratios at which bubble formation occurs and to characterize their evolution over time. We conducted microscopic experiments and numerical simulations of bubble dynamics at supersaturation ratios ranging from 0.1 to 3 on hydrophilic and hydrophobic surfaces. The results indicate that the initial bubble nuclei sizes are below 20 μm, significantly smaller than the conventional view of over 100 μm. The bubble growth behavior conforms to the Epstein-Plesset equation, and Ostwald ripening occurs at specific sites. Besides, surface wettability is proved to have significant influences on bubble nuclei size and density. These results provide experimental evidence that the existence of nanobubbles may contribute to inaccuracies in classical theoretical predictions.

The behavior of bubbles in multiphase flow systems is critical to many industrial applications, including nuclear reactors and separation processes. While significant research has been done on bubble dynamics, the effect of perforated plates in downward flow conditions remains less explored. This study aims to investigate the dynamics of air bubbles in a downward flow as they interact with perforated plates, focusing on bubble penetration probability and bubble residence time near the plate.

Preliminary results suggest that the perforated plate significantly interrupts the two-phase flow, with a smaller open area ratio leading to fewer bubbles passing through the plate. Experiments were conducted in a vertical water channel with a perforated plate positioned to obstruct the downward flow. To focus on the bubble motion, a single bubble was injected into the channel. The bubble motions were recorded using high-speed imaging, and varying flow rates and hole geometries were tested. The Bond number and other dimensionless parameters were analyzed to understand the characteristics of the flow and the air bubbles.

The study is expected to reveal how flow rate and perforation geometry influence bubble motion when the bubble encounters a perforated plate in downward flow. The findings from this research will contribute to a deeper understanding of bubble behavior in two-phase flow systems with perforated structures, which can inform the design of more efficient separation devices and reactors.

A series of experiments were conducted to investigate bubble characteristics in a vertical annular channel under a wide range of pressure conditions. For this purpose, a custom-designed 4-sensor optical fiber probe (4S-OFP) was developed to measure key local two-phase flow parameters, including local void fraction (α), bubble velocity (V_b), and Sauter mean diameter (D₃₂). The 4S-OFP demonstrated applicability in high-pressure, high-temperature environments up to 15 MPa and 350°C, with measurement uncertainties of approximately 2% for α, 5% for V_b, and 16% for D₃₂. The experimental conditions covered outlet pressures of 0.2–10.0 MPa, mass fluxes of 400–5,000 kg/m²s, inlet subcooling temperatures of 7–25°C, and heat fluxes of 300–620 kW/m². Local bubble parameters were systematically measured and analyzed under varying flow conditions, including mass flux, inlet subcooling, heat flux, and system pressure. The results showed that void fraction was strongly influenced by flow conditions, particularly heat flux and system pressure, while bubble velocity demonstrated a strong sensitivity to changes in mass flux. Additionally, the Sauter mean diameter (D₃₂) exhibited noticeable variations depending on both the inlet subcooling and the system pressure. These findings provide valuable experimental data for validating and improving existing thermal-hydraulic models, particularly under diverse pressure and subcooled boiling flow conditions. The dataset also highlights the importance of precise measurement techniques, such as the 4S-OFP, for advancing the understanding of local bubble dynamics in two-phase flow systems.

The objective of this study was to evaluate condensation heat flux q_c from steam and air mixture on a vertical flat plate, which is one of boundary conditions in CFD (computational fluid dynamics) analysis for thermal hydraulic behavior in a containment vessel during accident conditions. In our previous study, we carried out numerical simulation of wall condensation from saturated steam and air mixture on a vertical flat plate with the cooling height of 6 m by using the CFD code FLUENT, and evaluated effects of mixture velocity u_in on q_c. In this paper, we evaluated effects of steam mass fraction Y_s,in on q_c from saturated steam and air mixture on the vertical flat plate with the conditions of u_in = 0.53-3.2 m/s and Y_s,in = 0.226-0.68 by using the FLUENT code. The boundary condition for q_c, which was defined in the viscous sublayer, was used, and the size of the computational cell was 0.1 mm for the cells in contact with the condensation wall (where the dimensionless distance was y⁺ = 0.12-0.56). The q_c,CFD values computed with FLUENT were well expressed by an existing q_c correlation for forced convection (FC) condensation, but was a little larger than the q_c,cal values computed with an existing q_c correlation for natural convection (FC) condensation. The u_in value at the transition from FC to NC condensation became large for large Y_s,in due to large density difference.

The condenser is an important component in the secondary circuit of a nuclear power system. In marine applications, vibration and noise reduction are critical design goals of the condenser, with the heat transfer tube—its central element—playing a vital role in ensuring the reliability and longevity of the equipment. Two key physical phenomena occur within the condenser heat transfer tubes: condensation heat transfer and fluid-induced vibration. These phenomena are interdependent and continuously coupled. To investigate the vibration characteristics of heat transfer tubes during the condensation phase change, a visual experimental setup was designed. This setup allows for the observation of vibration behaviors and the analysis of how various parameters influence heat transfer performance. Modal tests revealed that temperature significantly impacts the natural frequency of the heat transfer tube, with the natural frequency decreasing as temperature increases. Dynamic tests demonstrated that the changes in volume and pressure due to condensation phase change are particularly significant at low steam flow rates. As steam velocity increases, the effect of condensation diminishes, and fluid shock becomes the dominant factor. Regarding heat transfer, an increase in the heat transfer rate leads to a higher vibration amplitude, while the vibration frequency decreases. These findings provide experimental and theoretical basis for optimizing condenser performance.

The application of a two-phase loop thermosiphon is evident in Passive Containment Cooling Systems designed to remove core decay heat following a Loss-of-Coolant Accident. It is characterized by two dominant effects, which are boiling and condensation. The phenomenon in practical applications usually includes the effects of noncondensable gas, however, only pure vapor cases are considered in the present study for simplicity.

The two fluid model (TFM) is widely applied in various code to analyze two-phase flow behavior in nuclear field. For the implementation of subcooled nucleate boiling, we use the RPI heat flux partitioning model in near wall region because it has been well described in literature and favoured in many modern CFD codes. Meanwhile, condensation rate in subcooled bulk is calculated by assuming zero resistance in vapor side via specifying an infinite value of vapor heat transfer coefficient. On the contrary to boiling model, only a limited number of film condensation models in TFM approach are present, especially for large-scale volume like the containment structure. Instead of resolving the thin film thickness, a subgrid liquid film model is implemented in wall-adjacent cells. Phase change rate at the cooling wall is computed by solving governing equations coupled with an additional mechanistic liquid momentum equation. In bulk region, a same approach as in boiling model is applied as we consider zero resistance condition for liquid side. Subsequently, the subcooled flow boiling and film condensation models are incorporated and specified in the corresponding domains to model a two-phase thermosiphon.

During a severe accident, the potential leaking of fission products (FPs) from a nuclear facility to the atmosphere represents a significant nuclear safety challenge. Accurate estimation of these releases is important to conduct an appropriate risk assessment and implement the necessary measures. To achieve this, IRSN conducts research dealing with the mitigation of FPs transported in a carrier gas injected through a liquid pool. This process, referred as 'pool scrubbing', can occur in several accident scenarios in light water reactors (PWRs, BWRs), including Filtered Containment Venting Systems (FCVS) or with the Steam Generator Tube Rupture (SGTR), as well as in nuclear-powered submarines or in new Small Modular Reactors (SMRs) using pressurised water.

Thus, experiments are currently being conducted to characterize bubble hydrodynamics and trapping of iodine compounds (decontamination factor measurements) on dedicated facilities, involving demineralized water and saline solution for different carrier gas injection flowrates at ambient conditions. In this context, advances results have been previously obtained on the hydrodynamics that occur along the pool [1] and on the retention of CsI aerosols [2] and volatile I₂. Building on this, a new study has been initiated to investigate the impact of a saline solution on these phenomena and to develop a more sophisticated bubble plume model.

First results suggest that saltwater generates smaller bubbles (except near the nozzle) and improves the retention of I₂ compared to clear water. Ultimately, these works will be used to enhance the pool scrubbing modelling implemented in the ASTEC integral code, developed by IRSN.

With the recent research on the miniaturization of nuclear power plants, it has become important to also miniaturize and modularize components, such as heat exchangers. Particularly, the steam generators used in newly developed SMRs (Small Modular Reactors) have adopted helical tubes instead of conventional U-tube designs to achieve compactness and enhance thermal efficiency. In PWRs (Pressurized Water Reactors), a loss of coolant accident (LOCA) leads to decrease the pressure and temperature of the primary side, causing the pressurized water to transform into two-phase steam. This steam flows downward from the upper shell side of the steam generator and condenses through the interaction with feedwater on the tube side. This study was conducted to evaluate the condensation heat transfer performance on the shell side under these conditions and to investigate the condensation heat transfer mechanisms. For this purpose, a test section was developed to assess the condensation heat transfer on the helical tube, and visualization experiments were performed to evaluate the behavior and thickness of the condensate film formed on the helical tube. Additionally, qualitative analysis of the condensation heat transfer mechanisms occurring on the helical tube was conducted based on the observed condensate film thickness from the visualization experiments and the measured heat transfer performance

Modern nuclear power plants often rely on passive systems, especially during severe accident scenarios. Systems like that utilize natural circulation and convection, which can challenge the capabilities of existing safety analysis codes. The forces governing the natural circulation phenomena are usually weak, and thereby, the natural circulation patterns can be easily disturbed. In accident analyses this may happen by physical phenomena or modeling/numerical issues. Therefore, the verification and validation (V&V) of these system codes are crucial for credible safety analysis to ensure the safety of current reactors and for licensing new designs, including Small Modular Reactors.

In this study, the authors analyze containment thermal-hydraulic phenomena using the MELCOR 2.2 code, drawing on selected experiments conducted at the Paul Scherrer Institute's PANDA facility. The focus is investigating natural convection and fluid circulation in containment-like geometries under accident conditions. As part of the preliminary analysis, the HYMERES HP6 experiments were chosen. In the tests, steam and He (to mimic H2) are injected into one of four PANDA vessels, to analyze circulation patterns between vessels, as well as atmosphere stratifications. The HP6 tests allow to explore various phenomena, including stratification, circulation, condensation, and the influence of non-condensable gases.

The results, which include sensitivity analyses, help identify containment phenomena that are well-represented by the MELCOR 2.2 code, as well as those that present challenges for accurate modeling. This work aims to contribute to potential future recommendations for improving the modeling of these complex phenomena.

The global energy landscape is undergoing a nuclear renaissance, driven by rising energy demands, economic growth, and the computational needs of artificial intelligence (AI) and machine learning (ML). With around 65 reactors under construction across 15 countries—and more planned, including in nations new to nuclear energy—the industry is advancing swiftly. While most new reactors rely on traditional light water reactor (LWR) designs, a growing number embrace advanced concepts like heat pipe, gas-cooled, liquid metal-cooled, and molten salt reactors. These “new” advanced designs, grounded in mid-20th-century thermohydraulic (TH) research, face modern demands for safety, economic viability, and regulatory compliance. A key challenge is the reliance on outdated TH experimental data, which, limited by past techniques’ resolution, lacks the spatial and temporal detail to capture next-generation reactors’ complex flow physics, hindering validation of simulation tools like CFD. In this paper, we present examples of high-resolution thermohydraulic experiments designed to support the development of advanced nuclear reactors. These experiments address key challenges in model validation by transforming raw measurements into trusted datasets. Using cutting-edge diagnostics and instrumentation, they deliver precise, high-fidelity data that capture critical flow phenomena under relevant conditions. The resulting datasets are used to validate CFD and multi-physics simulation codes, reduce modeling uncertainties, and support improved physical understanding. By strengthening the predictive capability of simulation tools, these experiments contribute to refined reactor designs, optimized performance, and more efficient regulatory licensing—ultimately enabling the safe and effective deployment of next-generation nuclear technologies.

Dispersed flow film boiling (DFFB) is a key flow regime that affects fuel rod integrity during emergency core cooling system (ECCS) injection phase in large-break loss of coolant accident in light water reactors. However, the existing experimental data in the DFFB regime have limitations in the important parameters that were measured, such as the void fraction and vapor superheat. These limitations in the DFFB data lead to remarkable uncertainties in the models/correlations developed for the wall convective heat transfer. To improve our understanding of and obtain experimental data for the wall heat transfer characteristics in the DFFB regime, a series of quasi-steady-state DFFB experiments was performed in the Post-CHF Heat Transfer (PCHT) test facility, over flow conditions of mass flux from 60 to 150 kg/m²-s, and pressure from 1.38 to 4.14 bar. The obtained wall convective heat transfer coefficient shows a transition in its trend with the thermodynamic equilibrium quality due to the transition of the dominant heat transfer mechanism from interfacial heat transfer to the vapor convection. The transition mechanism was confirmed by analyzing the computed vapor Reynolds number and the measured void fraction by an X‑ray radiography system. Based on the collected data, a new wall convective heat transfer correlation was developed by applying the Reynolds analogy to consider the variation of the interfacial area concentration with the droplet size. The newly proposed correlation successfully demonstrated its improved predictive capability compared to the existing models/correlations for a wide range of DFFB wall heat transfer data.

Best Estimate Plus Uncertainty (BEPU) methodologies are gradually establishing themselves as the favored way to perform Deterministic Safety Assessments (DSA) of Nuclear Power Plants (NPPs), because they account for uncertainties in plant states and physical behaviors while employing accurate-to-reality (best estimate) simulation codes. This work is performed in the context of the ATRIUM project, which seeks to establish best practices and standardize the BEPU process to make the results consistent, minimizing user effect. The approach taken in this project is to focus on specific phenomenology that is crucial to the progression of the desired transient (a Loss of Coolant Accident), and work with Separate Effect Tests (SET) experiments to find adequate uncertain parameters and their PDFs to propagate.

The target phenomenology for this study is that of post-Critical Heat Flux (CHF) heat transfer, concretely film boiling. Simulations of the Becker film boiling experiments were performed using RELAP5 to validate its suitability and identify influential parameters. Initial tests showed that RELAP5 overestimated burnout quality, resulting in inaccurate axial temperature profiles. Also, a sensitivity analysis of several RELAP parameters and initial conditions showed they didn’t influence the results.

To address this, RELAP5’s source code was modified to introduce and externalize new parameters for better control over film boiling modeling. Consequently, a sensitivity analysis was carried out on this new group of parameters, helping measure their effect and aiding in fine-tuning them to improve RELAP5’s predictions. This enhanced flexibility demonstrates a promising approach for more accurate RELAP5 analyses of complex post-CHF phenomena.

Prediction of the quench temperature is one of the most important parameters for the accurate estimation of the accident progress in light water reactors (LWR). In this study, a single rod flow quench experiments were conducted in the subcooling range of 0 to 40 K and 600-1100 °C initial cladding temperature conditions. Measurements of quench temperature were made at four levels of elevation from the bottom of the furnace. Based on these comprehensive tests, unique quench temperature trends were identified comprising two distinct regimes: (i) increase of quench temperature with increasing subcooling and initial cladding temperature with less effect of elevation in the high subcooling and low initial cladding temperature regime and (ii) constant quench temperature independent of subcooling and initial cladding temperature with a large effect of elevation in the low subcooling and high initial cladding temperature regime. The criteria delineating the two regimes was determined as a function of both subcooling and initial cladding temperature. A thorough analysis of cooling rate during the film boiling regime and quench front velocity was performed to develop a quench temperature correlation to better understand this new finding on quench temperature behavior. The study covers a critical gap in literature where high initial cladding temperatures (over 800 °C) under wide range of subcooling for such experiments have not been typically explored.

Quenching of hot vertical wall with a falling liquid film is an important thermal-hydraulic process to ensure the safety of nuclear reactors even during emergency situations. In this work, to develop a reliable model to predict the propagation velocity of the quench front, the temperature distribution of heat transfer surface during quenching was measured using a high-speed infrared camera. A silicon wafer that is transparent to the infrared ray was used as the hot wall, and the initial wall temperature, the wall thickness, the cooling liquid temperature, and the liquid flow rate were changed parametrically. The main heat transfer mechanism from the wall to the liquid film near the quench front was found to be the nucleate boiling. The heat transfer coefficient profile derived from the measured temperature distribution was therefore correlated using widely accepted heat transfer correlations such as Zuber's correlation for pool boiling CHF (Critical Heat Flux) and Rohsenow’s correlation for the pool boiling HTC (Heat Transfer Coefficient). The calculated quench velocity was in good agreement with the experimental data not only for the silicon wafer but also for the copper and zirconium walls of different thermal properties.

This study explores critical heat flux (CHF) in flow boiling on a downward-facing surface with non-uniform heat conductance properties. While prior investigations have examined non-uniform heat conductance surface, this research uniquely focuses on the downward-facing configuration, addressing an area that has been less explored. Additionally, the study expands the scope of surface modifications compared to earlier work, enabling a broader analysis of CHF behavior and non-unform heat conductance surface.

A significant advancement in this research is the introduction of a new CHF enhancement concept designed to prevent CHF initiation due to local hot spots. This concept is experimentally validated by measuring temperatures at both upstream and downstream regions of the heated surface, ensuring that local temperature variations and CHF dynamics are thoroughly understood. The methodology incorporates low thermal conductivity tape for surface modifications, chosen for its versatility and ease in creating various thermal conductance conditions.

Experimental results reveal a substantial CHF enhancement of up to 20%, highlighting the effectiveness of the proposed approach. These findings provide valuable insights into boiling heat transfer improvement strategies, particularly for downward-facing applications, and demonstrate the practicality of mitigating CHF triggers through innovative surface designs and precise thermal management.

Furthermore, the concept designed to prevent CHF initiation due to local hot spots can be applied to enhance the in-vessel retention capacity of a pressurized vessel. Since heat flux is non-uniform across different regions of the vessel, this approach could improve the overall CHF behavior, thus enhancing the thermal management and safety of pressurized vessels.

In this study, the effect of oxidation on the critical heat flux (CHF) of a vertical copper surface with four 16mm x 3mm x 2mm horizontal grooves and an bare copper surface will be investigated. During the experimental process, the copper surface will continuously oxidize with the repetition of experiments. For the bare copper surface, the oxidation rate is faster, as evidenced by its rapid darkening and loss of metallic luster, and the corresponding CHF value gradually increases with the surface oxidation. For the grooved copper surface, the oxidation rate of the surface is similar to that of the bare copper surface, but the oxidation rate within the grooves is much slower than that of the bare surface. The CHF value will remain stable within a certain range for a period of time before rapidly increasing to another level and continuing to remain stable. Overall, compared to the bare surface, the grooved copper surface has an enhancing effect on CHF, which is due to the change in macrostructure causing changes in bubble behavior, while oxidation also enhances the CHF of the copper surface, which is due to the change in the heating surface, thereby indirectly changing the bubble behavior.

Critical heat flux (CHF) prediction is essential for high heat systems like nuclear reactor and two-phase flow systems for enhancing dependability, safety and efficiency. CHF imposes design and operational restrictions due to safety concerns. However, there is work to be done to develop a reliable and effective CHF model. Machine learning techniques can find patterns and correlations in big datasets but lacks in explaining physical laws underlying CHF. Traditional ML operates as black-box which may result in physically unrealistic predictions, when applied to unforeseen circumstances and show instability. To address this, Physics-informed machine learning (PIML) integrates physical principles into PIML framework. While conventional ML uses only data, PIML integrates data-driven learning with knowledge from physical model. The goal of this study is to see the impacts of different physical models on reducing black-box nature of PIML and improving its interpretability for CHF prediction. In this study, four physical models for calculating CHF (Zuber, Katto-Ohno, Biasi and 2006 lookup table) were used as physical part of PIML and coupled with different pure MLs. A big amount of experimental data was used for the training and validation purpose of pure MLs and PIMLs. A thorough investigation has been carried out to assess (1) the predictive power of PIMLs and (2) compare the physical behaviors of mass flux, pressure, diameter, ratio of length-to-diameter and inlet subcooling on CHF from the PIMLs and pure MLs. This work shows the necessity of selecting a suitable physical model for approaching a robust and dependable PIML model.

Curved tubes such as U-bend, helical coil tubes etc., are commonly seen in the design of steam generators, heat exchangers in power plants due to their compactness and better heat transfer characteristics. Prediction of Critical heat flux (CHF) in the curved tubes is necessary in the system design for improved performance and safe operation. The occurrence of CHF is dictated by the geometric and operating conditions such as channel diameter, mass flux, subcooling, operating pressure etc. In the present study, a dimensionless number (De) determines the heat transfer and fluid flow characteristics in the curved geometries. The effect of De on the occurrence of the CHF in the curved tubes is given less attention in the literature. To this end, the simulations are performed using the two fluid model framework coupled with a wall heat flux partition (WHFP) model. The range of mass fluxes varies from 1500-3500 kgm^-2s^-1 and the degree of subcooling varies from 10-30 K . It was observed that the secondary flows created in the bent tube due to the centrifugal acceleration causes the fluid to undergo greater turbulence and flow separation in the bent region as De increases that enhances the heat transfer characteristics. A higher De causes the fluid velocity to increase which in turn causes the wall temperatures to drop. This further lowers the vapour volume fraction thus delaying the occurrence of CHF ensuring the safe operation.

Critical heat flux (CHF) is a primary power limiting criterion for water-cooled nuclear reactors, which must operate below the CHF conditions to allow sufficient thermal margin during normal operation. Accurate prediction of CHF is important for reactor design and safety analysis to determine safety margins under normal operating conditions, evaluate the maximum sheath temperatures of fuel bundles under anticipated operational occurrences, and predict the consequences under design basis accidents. To date, the most accurate CHF prediction method covering the widest range of flow conditions is the CHF look-up table. The latest version of the CHF look-up table was published in 2006. It was developed based on the world's largest CHF database as of 2005. Since then, a large number of CHF experimental studies have become available, allowing for the existing CHF prediction methods to be further improved. An upgraded CHF look-up table was derived from the expanded CHF databank containing 172 data sets with 42667 data points. The upgraded CHF look-up table provides good predictions for the range of flow conditions covering the normal operation, anticipated operational occurrences, and anticipated accident scenarios of water-cooled nuclear reactors. The upgraded CHF look-up table is expected to aid in improving subchannel and system thermalhydraulics analyses of reactor safety margins and consequences of postulated loss of coolant accidents for water-cooled reactors. The upgraded CHF look-up table also improves the accuracy of predictions under conditions related to fusion reactor divertor applications, and conditions corresponding to supercritical water reactor applications.

In the period of an intensifying energy crisis, the development of nuclear energy is of great importance. Fuel assemblies are critical components of reactor cores. As an innovative fuel type, Helical Cruciform Fuel (HCF) offers a larger heat transfer area per unit volume compared to traditional round fuel rods. Its unique helical structure enhances fluid flow and heat transfer capabilities. Additionally, the periodic contact formed by the helical structure provides self-supporting functionality, eliminating the need for position supporting and simplifying the structure of reactor core. These advantages make this novel fuel rod a focal area of research in Small Modular Reactor (SMR). In the course of extended operation of reactors, the pressure vessel may undergo deformation, causing displacement between the fuel rods and the pressure vessel and resulting in eccentricity. The heating curves of nuclear rod bundles typically exhibit non-uniform heating patterns in the reactor core. Unlike uniform heating methods, non-uniform heating introduces greater uncertainty in the location and values of critical heat flux (CHF). In this paper the RPI boiling model combined with the Eulerian-Eulerian two-fluid model are used to investigate the subcooled boiling and critical heat flux (CHF) heat transfer characteristics of HCF under non-uniform and uniform heating. Besides the influence mechanisms of different eccentricities on HCF with non-uniform and uniform heating power curves are explored. The findings of this paper will provide valuable insights for further research and practical applications of HCF.

In order to enhance the safety of nuclear power plants, it is required to establish emergency cooling methods for reactor accidents. In PWR, the using In Vessel Retention (IVR) method is considered to prevent melt-through in meltdown. In the IVR method, the cavity surrounding the reactor pressure vessel is filled with water in the IVR method to enable cooling by boiling heat transfer. The maximum cooling capacity of boiling heat transfer is determined by the critical heat flux (CHF), and improving CHF is crucial to implement the IVR. In this study, we found that forming a porous copper structure on the boiling surface improved the CHF to approximately 5 MW/m², which is about four times higher than that of an uncoated plain surface. The boiling experiments were conducted using a heat transfer surface with a diameter of 10 mm and porous copper structures with thicknesses from 0.5 mm to 3.4 mm under saturated temperature conditions at atmospheric pressure. It was observed that CHF increased as the thickness of the porous structure increased up to 2 mm, but decreased when the thickness reached 3.4 mm. In this presentation, we will discuss the factors contributing to the CHF improvement.

In the face of the escalating energy crisis, the advancement of nuclear energy assumes paramount significance. Fuel assemblies, crucial elements of reactor cores, play a pivotal role in this domain. During the long-term operation of reactors, deformation of the pressure vessel may occur, leading to displacement between the fuel rods and the pressure vessel, resulting in eccentricity. Studying the subcooled boiling and critical heat flux (CHF) phenomena within annular pipes and exploring the effects of eccentricity are crucial for clarifying the two-phase boiling processes in fuel assemblies and enhancing reactor safety. In the reactor core, usually the heating curves of rod bundles are non-uniform heating. Comparing to the uniform heating methods, the non-uniform heating mothods bring more uncertainness in CHF locations and values. Using CFD analysis software, in this paper the RPI boiling model combined with the Eulerian-Eulerian two-fluid model are employed to analyze the subcooled boiling and CHF heat transfer characteristics of annular pipes under non-unifrom and uniform heating. On this basis, the mechanisms underlying the effects of different eccentricities on annular pipes with non-uniform and uniform heating are explored. The findings of this paper provide valuable insights for a deeper understanding of heat transfer phenomena within fuel assemblies and offer guidance for their practical application.

During the operation of nuclear reactors, corrosion particles deposited on fuel cladding alter its surface characteristics, thereby influencing the critical heat flux (CHF). The modification of surface properties, such as roughness, wettability, and porosity, plays a significant role in determining the heat transfer efficiency and safety margins of the reactor. However, the absence of detailed structural parameters for the deposition layer has hindered a comprehensive theoretical analysis of the mechanism by which the deposition layer affects CHF. To address this gap, the present study conducted systematic pool boiling deposition experiments under varying deposition times and heat fluxes. The experimental results demonstrated that the average CHF of smooth rods was 1240 kW/m², with a deviation of less than 10% from model predictions, thereby confirming the measurement accuracy and stability of the experimental apparatus. This validation establishes a robust and reliable experimental platform for subsequent research on CHF behavior under deposition conditions. Notably, the average CHF of rods with deposition reached 1479 kW/m², representing a 19.2% enhancement compared to smooth rods. This significant improvement suggests that, in terms of CHF enhancement for fuel rods, corrosion particle deposition exerts a positive influence. These findings contribute to a deeper understanding of the complex interactions between surface properties and CHF, offering critical insights into the role of deposition layers in enhancing thermal performance.

Enhancing our understanding of the link between the near-wall void-fraction profile and nucleation characteristics at the boiling surface in subcooled-flow conditions is crucial for improving subcooled-flow and DNB models. This interaction is being investigated at a flow-boiling facility at MIT, which features a deionized water loop capable of operating at pressures up to 10 bars and mass fluxes up to 2000 kg/m²/s. The test section includes a rectangular flow channel (3 cm x 1 cm).

Subcooled nucleate boiling was generated using a custom-made heater, consisting of a sapphire substrate coated with a thin layer of chromium, housed in a heating cartridge mounted on one of the test section walls. An optical probe, translated with sub-millimeter accuracy to and from the heated surface, measured the void-fraction profile near the surface. Infrared measurements through the back of a sapphire substrate allowed for the quantification of wall temperature, heat flux, and relevant nucleation properties at several heat fluxes.

The collected measurements are the first step towards forming a comprehensive picture to elucidate the mechanisms linking the void fraction distribution in the vicinity of the boiling surface with the boiling process and the boiling crisis. Insights gained from this study may inform the development and validation of next-generation models for flow boiling simulations.

To prevent reactor vessel failure, the in-vessel corium retention through the external reactor vessel cooling (IVR-ERVC) has been adoptedas the severe accident mitigation strategy in nuclear reactors. In existing large nuclear reactor types, IVR-ERVC is conducted in a natural flow condition, which forms between the insulator and the reactor cavity. However, in Small Modular Reactors (SMRs) currently under development in the Republic of Korea, there is no insulator due to the integral design, and thus IVR-ERVC is conducted in a pool condition. Additionally, since the reactor lower head outer wall of SMRs is made of stainless steel, it is essential to study the Critical Heat Flux (CHF) phenomena occurring on stainless steel surfaces to ensure the safety and effectiveness of the IVR-ERVC process in these reactors. In this study, focusing on the integral SMR design, the experiment measured CHF under various conditions, including the effects of heater surface inclination and material properties for stainless steel. Experiments using an SS304 heater in pool boiling conditions are conducted to develop a modified CHF correlation, reflecting the specific characteristics of integral SMR, challenging existing models and contributing to safer nuclear power technology.

The critical heat flux (CHF) is a key thermal limit in water-cooled nuclear reactors and accurate and reliable modelling of boiling and CHF remains an unresolved challenge in nuclear thermal hydraulics. In large majority, CHF is still estimated using empirical models derived from expensive, full-scale experiments. Due to the empirical nature of the models, significant engineering margins are applied, restricting reactors to operate at a power level that is below their theoretical potential.

In the last few decades, computational fluid dynamics (CFD) models based on the multifluid Eulerian-Eulerian method and the heat flux partitioning approach have shown promise in reducing conservative design margins through more accurate predictions. However, the large number of modelling closures required (e.g., nucleation site density, bubble departure diameter) and the overfitting of the numerous constants on limited datasets has so far prevented developing a universally accepted, best-possible model and deliver the anticipated improvements.

In the last few years, advancements in measuring techniques have made possible detailed, small-scale measurements that enable the validation of boiling models at a level of detail that was not possible before. In this work, we have developed and implemented in MATLAB a heat flux partitioning framework and, leveraging these new data, assessed the most frequently used and recent heat flux partitioning models in pool and flow boiling conditions. Strengths and weaknesses of each model, and some physical inconsistencies are identified. Impact of uncertainty in closure models is quantified, improvements implemented and validated and areas for future development suggested.

Flow-induced vibrations (FIV) of key nuclear reactor components, such as fuel rods and steam generator (SG) tubes, may lead to wear and damage. Hence understanding and being able to predict the vibrational behavior of these components is crucial to mitigating any potential risks and preventing undesired outages. Fuel rods primarily vibrate due to the turbulent axial flow, requiring generally scale-resolving models to properly study their vibrations numerically. SG tubes on the other hand are exposed to cross-flow, with vibrations being a result of a combination of turbulence-induced and vortex-induced vibrations, possibly resulting in fluid-elastic instability. This cross-flow nature of the problem may make it possible to study it using computationally less expensive numerical techniques, such as those based on the Unsteady Reynolds-Averaged Navier-Stokes (URANS) approach or hybrid turbulence models. The current paper attempts to give an overview of where we are in terms of our understanding of FIV of a multiple tube configuration in cross-flow and how well these problems can be modelled using medium-resolution numerical approaches. This is done by first considering two well studied problems, being the numerical benchmark of Turek & Hron of a flexible flap attached to a fixed cylinder, and a single cylinder in cross-flow. The former allows one to validate properly the FSI framework used to study cylinders subjected to cross-flow, while the latter serves as a canonical problem fundamental to understanding tube bundles in cross-flow. Following these two cases, two-cylinder systems, with cylinders positioned either inline or side-by-side, and tube bundles are discussed. In general, a lot of data coming from experiments is available for all these cases, allowing one to validate and study them numerically in detail. Also, medium-resolution simulations do provide reasonable predictions for single and two-cylinder configurations, but struggle to recover vibration amplitudes in the lock-in regime. For tube bundles though, the amplitudes are generally overpredicted. This may be caused by a lack of turbulence that is actually resolved, although more detailed benchmark data is needed to further investigate this.

Improving thermal-hydraulic performance is a major goal for many applications since fluid flow is essential to many natural and artificial systems. This study focuses on assessing thermal and fluid flow performance in a corrugated mini channel, which has a distinct separation and mixing zone arrangement that influences its thermal-hydraulic behavior. To investigate how various geometric parameters affect this channel's hydraulic performance, experiments and CFD simulations were carried out. Using water as the working fluid and volumetric flow rates ranging from 1 to 7 L/min, increasing in increments of 0 to 1 L/min, an experimental investigation was carried out. The Reynolds numbers for these flow rates ranged from 1000 to 4000. The study also explores the effect of the mixing-to-separation-zone length ratio (Lm/Ls) on hydraulic operations. A crucial metric for evaluating hydraulic performance, the friction factor, and Lm/Ls are clearly correlated in the experimental results. This experimental result had a maximum deviation of 5% from the numerical calculation. Consequently, a power law-based novel correlation with a variance of less than 5% is suggested to forecast the friction factor and heat transfer. This emphasizes how the Reynolds number and geometric parameters both affect the friction factor, a crucial hydraulic performance metric.

The accurate prediction of two-phase flow void fraction and wall superheat under pressurized conditions is crucial for understanding reactor safety margins. Existing void fraction and flow boiling heat transfer models used in numerical simulations exhibit significant uncertainty, limiting their accuracy in two-phase CFD simulations. This paper presents high-resolution vertical upward flow boiling experimental data from the PCHT test facility at the University of Michigan, using a gamma densitometer and X-ray imaging system. Experimental results are compared with established one-dimensional models to validate their applicability and identify limitations. The Saha and Zuber correlation is used to predict the thermodynamic equilibrium quality at the net vapor generation point. The slip ratio model proposed by Chisholm, Thom, Zivi, Cai, Lockhart, and Martinelli was used to estimate the void fraction and heat transfer coefficient in sub-cooled flow boiling conditions.

Hence, based on the existing correlations from previous research and experiment data from the PCHT test facility, a new one-dimensional model is proposed to better predict the void fraction and wall heat transfer coefficient under the vertical, upward, sub-cooled flow boiling conditions.

This study presents an innovative use of boiling acoustics techniques to examine the cooling system of the Isotope Production Facility (IPF) at Los Alamos National Laboratory (LANL). In the IPF target station, multiple stacked targets are arranged with a series of water channels interspersed in between, leveraging forced convection for cooling. During operation, the rastered high-energy proton beam can initiate various boiling regions from subcooled boiling to critical heat flux. Identifying these boiling characteristics is challenging because of the extreme radiation environment. For that, a prototypical facility setup with a transparent window is utilized for the visualization of boiling phenomena. In this paper, we employ a hydrophone and high-speed video camera to capture acoustic signals and images indicative of various boiling phenomena. By applying signal processing techniques such as Fast Fourier Transform (FFT) and Short-Time Fourier Transform (STFT), we aim to discern distinct boiling behaviors from the hydrophone data. The insights gained from this analysis will guide the installation of hydrophones within IPF, allowing real-time monitoring to prevent boiling crisis by adjusting operational parameters such as beam intensity. While this methodology is tailored for IPF, its implications extend to other systems where boiling dynamics are critical, particularly in the nuclear industry and research sectors. This research enhances our understanding of thermal hydraulics and heat transfer in isotope production facilities and contributes to the safety and efficiency of nuclear systems.

This study investigates the nucleation site density (NSD) on nuclear fuel cladding materials under flow boiling conditions in an annular gap geometry. Experiments were conducted at the Karlsruhe Institute of Technology (KIT) COSMOS-L facility using three cladding samples: uncoated Zircaloy-4, and two physical vapor deposition (PVD)-coated variants, CrN and Cr. The test section comprised a 9.5 mm diameter cladding heated over a 330 mm length, with data collection focused on a 25 mm segment near the outlet. Measurements were performed at 300 kPa outlet pressure, approx 500 kg/m^2/s mass flux, and an 85°C inlet temperature, with variable heat flux. High-speed videography captured bubble dynamics, and nucleation sites were identified using an in-house KIT code that tracks brightness changes in individual frames to calculate the frequency and spatial distribution of departing bubbles. To distinguish true nucleation sites from passing bubbles, noise, and non-uniform illumination, an adaptive filter based on proper orthogonal decomposition was implemented. The NSD comparison across the three samples revealed observable variations, which require further evaluation under different flow and heat flux conditions to better understand surface modification effects on boiling behavior.

An experimental setup has been designed and manufactured at the Royal Institute of Technology (KTH) to investigate the internal structure of boiling two-phase water flow under prototypical Light Water Reactor core conditions, including those relevant to PWR, BWR and SMR designs. The setup is based on the High-pressure WAter Test (HWAT) loop, designed for 25 MPa pressure, 1 kg/s water mass flow rate and 1 MW thermal power. The facility has been updated with a new test section and advanced instrumentation systems to enable measurements under steady-state and transient operations. This novel experimental setup allows for the first-time measurements of radial distributions of local two-phase flow parameters under high-pressure LWR core conditions. The resulting data is intended to enhance the fundamental understanding of boiling two-phase flow phenomena, contribute to the development of closure laws and support the validation of computational codes. The paper presents the loop design, the updated instrumentation with associated uncertainties, and data post-processing methods (including the derivation of dispersed phase length scales). Results from commissioning tests, such as heat balance tests and single-phase tests, are presented. Examples of high-pressure boiling two-phase flow measurements are presented and discussed. Fundamental behavior and associated key parameters, including radial distributions of void fraction, mixture velocity, interfacial length scales and polydispersed characteristics, are identified and quantified.

The heat transfer mechanisms of flow boiling are still unclear. Recent research shows that evaporation of microlayer contributes to bubble growth in pool boiling. In order to investigate the microlayer heat transfer and dynamics, the 650nm-thick indium-tin-oxide (ITO) film is deposited on a 1mm-thick sapphire substrate to heat the fluid. The thickness of the microlayer underneath the bubble was measured using high-speed laser interferometry (LIF) and the transient temperature distribution on the wall was measured by an infrared (IR) camera. Another camera was used to capture the side bubble image. The three devices worked synchronously. The experiment was conducted at 0.11MPa with deionized water as the fluid, covering heat fluxes of 110-174.4 kW/m², subcooling degrees of 0-11.2 °C, and liquid flow velocity of 0.12-0.27 m/s. The typical bubble behavior was analyzed, including bubble growth, sliding, departure and wall temperature distribution. The inverse heat conduction problem (IHCP) of the boiling surface was solved based on conjugate gradient method (CGM) for wall heat flux partitioning. The wall heat flux partitioning outcome has revealed that the formation and evaporation of microlayer have an important effect on the growth of flow boiling bubbles.

Annular two-phase flow occurs in multiple types of nuclear reactors under normal operating conditions and accident scenarios. Annular flow typically features high flow quality and relatively thin liquid film around the fuel elements in nuclear reactors. This flow structure has significant safety implications since the liquid film could break into rivulets due to vaporization and entrainment. Important parameters to characterize annular flow consist of liquid film thickness, wave velocity on the surface of the liquid film, the size and velocities of entrained liquid droplets, and entrainment fraction and rate. This study contributes to existing database of annular flow with both liquid film and droplet measurements conducted with an air-water test facility. The experimental test section consists of a vertical pipe with an inner diameter of 9.525 mm and a length of 2.9 meters, as well as two measurement ports designed for the measurement of liquid film thickness and surface wave velocity by two sets of parallel-wire conductance probes placed at each port. To capture liquid droplet size and velocity, two high-speed cameras are used to capture the droplet field as they exit the test section outlet after extraction of liquid film. The annular flow testing matrix consists of an array of inlet conditions with superficial gas velocity ranging from 7.80 m/s to 34.91 m/s and superficial liquid velocity ranging from 0.09 m/s to 0.44 m/s, spanning the entrainment and non-entrainment annular flow regimes. Acquired data has been used to validate and improve existing correlations or closure models for annular flow.

The experimental characterization of thin liquid films in annular flow regime is central in many engineering applications, ranging from chemical industry to refrigeration systems, and in particular to cooling of light water nuclear reactors, where it is crucial for safety thermal analysis of boiling water reactors and for validation of system codes as well as CFD codes. However, characterization of thin films in annular flow is particularly challenging given the flow turbulence and the high non-linearity of the free-surface behaviour. Particularly, film thickness and wave characteristics are very challenging to be measured, with many correlations from the literature showing substantial offsets in predicting the same quantities under seemingly close boundary conditions. In this paper, results of simultaneous measurements of vertical upward annular flow film in an adiabatic test section is presented using three different techniques. These consist in: a) total internal reflection method, providing highly resolved local thickness measurements; b) X-ray attenuation method, providing interfacial topology and void fraction that can be converted into thickness information; and c) a conductivity film sensor providing high speed thickness and wave information with a spatial resolution of 2 mm. By performing independent calibrations, the three techniques are cross-validated within the corresponding uncertainties. To the authors’ knowledge, this is the first time that the three measurement techniques for film thickness are combined, thus constituting a unique benchmark. The three techniques complement each other and provide highly reliable measurements of annular flows, which are also compared to existing correlations available in the open literature.

Safety evaluation of offshore floating nuclear plants requires boiling two-phase flow data under heaving conditions with long wave period (10–20s). Void feedback effect necessitates investigating the mutual influence of oscillating coolant flow and thermal power.

This study conducted forced convection boiling experiments with sinusoidally oscillating coolant flow velocity and thermal power in an annular double wall channel with an electrically heated rod at atmospheric pressure. The inner diameter of flow channel was 20 mm, and outer diameter of the heater rod was 10 mm, with a heated length of 2000 mm. The axial power profile of heater rod was uniform. The time-averaged inlet flow velocity and linear power density were maintained 2 m/s and 15.7 kW/m respectively. Experiments considered no oscillation and 0.05 and 0.1 Hz sinusoidally oscillating inlet flow velocities and thermal powers with ±15% zero-peak amplitude. The distributions of two-phase flow parameters and flow regimes were identified using an optical void probe with radial traverse at two heights and stereo high-speed camera. An optical fiber sensor with a sheath tube was mounted on the heater rod surface to obtain the axial temperature distribution and capture heat transfer characteristics.

The void fraction oscillated corresponding to the inlet oscillations, significantly at the top of the heated area and slightly in the middle, and its amplitude was affected by the oscillation frequency. This behavior may be attributed to the oscillatory acceleration of the fluid flow. The radial distribution showed less bubble formation in the middle rather than at the top.

The experimental characterization of thin liquid films in multiphase annular flows is particularly relevant to the safety analysis of the cooling channels of light water reactors and for validation of best-estimate system codes as well as CFD codes. The total internal reflection method (TIRM) is an optical method known for decades for being able to non-intrusively measure film thickness of a wide range of fluids flowing over a transparent wall. This measurement is performed by recording with a camera the reflected circular pattern of a laser beam pointed to the flow. The transparent wall is often curved (such as in a pipe), which leads to a potential loss of information, since part of the reflected pattern has to be discarded from each frame because of optical distortions from the curved wall surface. This is also the case for the TIRM experiments performed in our laboratory on adiabatic vertical upward annular flows in a circular section pipe. However, in this work, an innovative approach is developed to use the information on the shape of the distortion rather than discarding it, thus maximizing the value of the measurement and improving accuracy. This is achieved thanks to a thorough data analysis backed up by a previously validated optical ray tracing simulation that replicates our TIRM experiments including the distorted patterns. From the combination of simulation and experiments, new insights are gained into the potential and the limits of standard TIRM film thickness measurements applied to curved pipes.

A new experimental setup has been constructed for a cold swirling turbulent jet issued through a pressure-swirl atomizer generating a spray at Reynolds number up to Re=10⁶.

The cold spray injection is used for steam condensation and pressure regulation in the pressurizer of a pressurized water reactor (PWR).

For large spray flowrates the droplets also reach the pressurizer tank wall, which acts as an undesired thermal load.

Current simplified prediction tools for transient load calculations lead to conservative estimations of the loads and under predicted lifetime.

More advanced tools, e.g. computational fluid dynamics (CFD) require better models for two-phase flow heat transfer in order to get more reliable lifetime predictions in a long term operation (LTO) context.

Measurements have been conducted in a 1:1.84 lab scale model of the spray two-phase flow characterizing the liquid fraction, droplet size and velocity distributions dependence on the spray flow rate and surface tension.

The spray cooling heat transfer has also been measured using a unique heat transfer sensor developed at CEA in France.

The experimental data base will be used for validation of more advance CFD models, being developed in conjunction to the present experimental campaign.

With the development of surface engineering, porous surfaces have emerged as a significant research subject in boiling heat transfer. The latter, in turn, plays a crucial role in various industries such as power plants, distillation plants, and microelectronic technology. In this paper, the Molecular Dynamics method is adopted to investigate the wicking dynamics and boiling dynamics of two porous surfaces: foam, which exhibits randomly distributed pores, and mesh, composed of ordered square wires with relatively uniform pore sizes. Three wettability, namely hydrophilic, neutral, and hydrophobic wetting states, are assigned to the two porous surfaces de-coupling the effect of wettability from surface structure. Results reveal that, during the wicking process, the foam surface shows better wetting ability as it absorbs liquid under both hydrophilic and neutral wettability. Comparatively, the mesh surface has the fastest wicking speed under hydrophilic wettability yet it becomes non-wetting under neutral wettability. During the boiling process, the boiling dynamics differ greatly under three wettability. More importantly, the difference in surface structure makes the foam surface possess a better heat transfer whereas the mesh surface causes gentle pressure variation. Our findings provide insights into the design of artificial porous surfaces for certain purpose and their potential application.

Water entrainment at T-junctions is of upmost importance in some nuclear safety analyses. Such a phenomenon directly impacts the core liquid inventory, hence its coolability. Numerical modelling using system-scale codes is essential for characterizing the two-phase flow interactions at the T-junction and in the branch line upstream. In light of this, code validation must be carried out through comparison to experimental data (Separate Effect Tests).

The test section represents, at a lower scale compared to reactor scale, an upper core plenum, a hot leg, and a pressuriser surge line. The test loop is operated at atmospheric conditions. Thus, the transposition issue (geometry and thermal hydraulics conditions) also has to be tackled.

The aim of this paper is to present the set of calculations, the comparison to experimental data and the scaling approach through the confrontation of CFD and system-scale code predictions.

The system-scale computations are performed with the thermal hydraulics code CATHARE and the CFD calculations with NEPTUNE_CFD, an in-house code.

CATHARE and NEPTUNE_CFD results are first compared to the experimental data, both, qualitatively (video recording) and quantitatively (water height) for two configurations (vertical upward and inclined T-junctions). This allows an assessment of the codes’ accuracy regarding the phenomenon of water entrainment at a T-junction and raises reflections on the physics, and the modelling.

Then, CATHARE and NEPTUNE_CFD calculations are performed at reactor scale and thermal hydraulics conditions. It is assumed that CFD better copes with scaling and is taken as a reference for the code-to-code comparison.

Finally, future work is proposed.

The efficient heat dissipation of the permanent magnet propulsion motor rotor is crucial to the development of advanced propulsion systems. As an advanced thermal management technology, the rotating heat pipe enables effective cooling of rotating components through internal phase-change heat transfer and natural circulation. Based on this research background, our team has built a rotating heat pipe experimental system and conducted experiments with a 70% filling ratio, a length of 500mm, and a diameter of 30mm using a stepped rotating heat pipe. The results show that its heat transfer capability gradually increases with the rotational speed. Under the same conditions, the heat transfer performance of the parallel-axis rotating heat pipe reached twice that of the coaxial rotating heat pipe, with an equivalent thermal conductivity of up to 1438.08 W/(m·K). This study provides experimental data support for the application of rotating heat pipes in the cooling of permanent magnet propulsion motor rotors.

The pool boiling experiment for the observation of corrosion products deposition is carried out to better understand the fouling mechanism under nucleate boiling conditions. The experimental apparatus comprises the quartz glass cavity, test piece (aluminum), high-speed camera and heater set-up. The deposition tests are performed in dilute colloidal solution (Fe₃O₄) with different wall temperature and bulk temperature at atmospheric pressure. The experimental observations indicate that the deposits exhibit a circular distribution and a thickness of approximately a few micrometers under nucleate boiling. The fouling ring is distinguished by a lower central thickness and a higher edge thickness. To gain further insight into the flow field distribution during the bubble growth process, the numerical simulation of the bubble growth and detachment process is conducted using the CFD method. It has been demonstrated that corrosion products are transported to the contact line of the bubble as a consequence of turbulence vortex. Besides, the micro-layer situated at the base of the bubble will undergo a process from thinning to drying out, resulting in the deposition of corrosion products on the heated surface. Through a combination of experimental and numerical techniques, the transport mechanism of corrosion products under the influence of nucleate bubbles has been elucidated, and the model between the evaporation flux and deposition rate of corrosion products has been developed.

The Best Estimate Plus Uncertainty (BEPU) methodology, developed over several decades, has seen numerous innovations aimed at enhancing the efficiency and quality of the BEPU procedure. Wilks’ formula characterized by nonparametric statistics is widely used for uncertainty evaluation, while it is time consuming. Deterministic sampling (DS) methodology assesses the uncertainty of outputs through the first two orders of moments of the input uncertain parameters. The reduction in computational effort achieved by using fewer sampling times, compared to the Wilks method, presents a promising alternative for enhancing the BEPU methodology. 16 input parameters and 3 safety-related output parameters as the Figure of Merits (FoMs) are chosen in ESBWR initiated by main steamline break for BEPU evaluation using RELAP5. First order Wilks’ method and three DS (DS-Standard, DS-Simplex, and DS-Hadamard) methods are applied. Subsequently, a preliminary sensitivity analysis of the Wilks’ results is performed to identify the input parameters with a significant impact on the FoMs. The downscaled parameters were then used as inputs for BEPU calculations using three DS methods. The degree of envelopment and conservatism of the three results (Wilks’ results with 16 input parameters, three DS results with 16 input parameters, and three DS results with downscaled parameters) relative to the experimental data were compared to determine whether the downscaled input results could be considered valid for the final BEPU analysis under the given conditions.

Nuclear energy provides about 70% of France’s electricity, with 56 pressurized water reactors (PWRs) operated by Electricité de France (EDF). EDF R&D uses advanced fluid mechanics to ensure reactor safety, employing in-house 3D codes like neptune_cfd to study two-phase flows and critical phenomena such as the boiling crisis.

This article focuses on phase change at the interface between liquid and vapour, often referred to as bulk or interfacial condensation/boiling. Although overshadowed by wall-driven condensation/boiling, interfacial phase change is crucial in nuclear applications, particularly in liquid metal flows for sodium-cooled reactors and microchannels used in the nuclear industry.

While single-fluid Volume of Fluid (VOF) codes effectively model interfacial phase change, multi-field computational fluid dynamics (CFD) codes lag behind. This article introduces a new phase change model for multi-field codes, using the gradient method to capture interfacial phase change accurately. The model is validated against both analytical and experimental cases involving bulk boiling, showing excellent agreement. Its mesh convergence aligns with single-fluid codes, and we propose a hybrid approach combining this model’s accuracy with the computational efficiency of dispersed-phase methods for simulating complex two-phase flows.

A novel analytical solution-based thermal resistance network computational method has been proposed to provide a more accurate and reasonable temperature calculation framework for pebble bed reactors. This method generates positional and contact information for large-scale pebble beds based on the analytical solution of multidimensional generalized thermal resistance and results from the discrete element method. It also calculates the generalized thermal resistance between the center points of adjacent particle contact surfaces. The generated data is trained using decision tree and random forest algorithms, constructing multiple weak classifiers (i.e., individual decision trees) and combining them into a strong classifier to reduce overfitting and enhance the model's generalization capability. A random forest model was built on the TreeBagger framework, utilizing 100 decision trees and 3 leaf nodes. The importance of each feature's impact on thermal resistance was analyzed, and the trained values effectively reflected the thermal resistance values, achieving a maximum percentage error of 1.45% in the testing set. Validation was conducted using simple cubic, body-centered cubic, and face-centered cubic packing arrangements, showing good agreement with finite volume method results. The novel thermal resistance network model was applied to compute the temperature field caused by heat conduction in the HTTU, confirming the model's feasibility and providing the temperature distribution at various nodes.

The accumulation of Corrosion-Related Unidentified Deposit (CRUD) in the core of the Pressurized Water Reactor (PWR) poses potential threats to reactor safety. This study investigates the deposition behavior of CRUD on the PWR fuel cladding surface, constructing a high-precision 3D CRUD dynamic growth model within a complex coupling framework. The Discrete Phase Model (DPM) is employed to analyze the transport and deposition processes of particulate corrosion products within the fuel rod bundle channels. By integrating the Zeta potential and Electrical Double Layer (EDL) model, the study systematically examines how the Zeta potential near the fuel cladding influences particle deposition behavior. Dynamic mesh technology is used to visualize the dynamic growth process of CRUD layer. In parallel, species transport equations are employed to analyze the distribution characteristics of metal ion concentrations and pH within the coolant. Finally, a multi-physics coupling mechanism of CRUD deposition behavior with rods channel inclusion, water chemistry, boiling heat transfer, and flow field distribution is revealed. The results show that Zeta potential and local pH affect the deposition behavior of corrosion product particles. Specifically, when the Zeta potential of the fuel cladding wall is positive, deposition becomes more challenging in acidic conditions but easier in alkaline environments as the Zeta potential increases. Furthermore, the rough surface of the CRUD layer induces localized accelerated flow near the cladding wall, which exacerbates electrochemical corrosion. The complex coupling effects of CRUD layer thickness and temperature field, particle deposition behavior with Zeta potential, local accelerated flow and electrochemical corrosion are revealed.

Flows past a solid wall for a well-known region between the wall and the bulk of the flow, the boundary region. The characteristic thickness of this boundary region is defined by the appropriate diffusion coefficients and is a place of high gradients and non-linear behaviour. In simulations, we prefer to avoid the calculation of the flow properties in the inner boundary region, since it increases the computational cost of the simulation greatly, being a very thin region next to the walls. This is possible, as the inner boundary layer exhibits a self-similar behaviour, that can be described with explicit functions, called wall- functions.

For thermal fields, the temperature gradient at the wall-normal direction determines the heat extracted from the wall, therefore its correct representation will determine the overall temperature field in the domain. It is therefore important we accurately compensate for the effect of the wall on the rest of the flow, if not resolved.

In the proposed paper we examine a turbulent boundary layer with a DNS with multiple temperature fields of various Prandtl numbers to design more accurate thermal wall-functions. The simulations are performed by the incompressible Navier-Stokes solver Nek5000 and restricted to forced convection flows. We will use these simulations to establish highly accurate explicit wall function that depends on the Reynolds and Prandtl number, making it applicable for a wider range of fluid flows.

This paper presents the 3D modelling and CFD analysis of a CANDU irradiated fuel bay (IFB). CANDU IFBs are significantly different from light water reactor spent fuel pools in terms of the bundle type and orientation of the assemblies. Following the Fukushima Daiichi accident, the Canadian Nuclear Safety Commission (Canadian regulator) undertook a vigorous re-evaluation of the current safety measures and margins of the CANDU IFBs under various stages of an extreme beyond design basis accident scenario. Amongst a few phenomena and hypothetical scenarios of interest, advancing the knowledge of the air-cooling effect on the fuel assemblies and storage racks during the complete loss-of-coolant accident (LOCA) was deemed significant. Under the pan-Canadian PIRT effort, the air-cooling effect on the fuel assemblies, which is driven by the natural convection and radiation modes of heat transfer, was identified as an area of high-importance with low knowledge-level phenomenon for IFBs. The objective of this study is to simulate the airflow and temperature distribution around the irradiated fuel racks at various decay power under a postulated accident scenario (complete LOCA) using the CFD code Simcenter STAR‑CCM+. A detailed 3D model based on the geometry of the CANDU fuel storage module was developed that was qualitatively analyzed in CFD for its capabilities to predict the sheath temperature of the irradiated fuel; a key parameter to monitor the severity of the IFB LOCA. It is anticipated that the developed CFD model could be leveraged to inform lower-fidelity codes and to guide experiments to develop validation data.