Conference Agenda

In water-cooled nuclear reactors, restoring core cooling after a Loss-of-Coolant Accident (LOCA) is critical, typically achieved through reflooding. During this process, the peak cladding temperature (PCT) arises between the dispersed flow film boiling and high-temperature vapor flow stages. Accurate PCT prediction is vital for reactor safety, spurring extensive research into droplet-wall heat transfer during high-temperature collisions. The boiling regimes—distinctive heat transfer mechanisms determined by wall temperature—necessitate precise identification for reliable modeling. However, previous studies using either hydrodynamic or thermal visualization to classify boiling regimes often yield inconsistent criteria due to their reliance on single techniques. This study addresses these limitations by simultaneously capturing hydrodynamic behavior and thermal characteristics, enabling improved boiling regime identification and comprehensive analysis of heat transfer mechanisms. Experiments used a circular substrate with two visual fields: a transparent section for observing droplet dynamics and an infrared-opaque section for thermal footprint detection. Substrate temperatures ranged from 150°C to 600°C, with droplets at saturation temperature released under gravity at a Weber number of 50. Side-view imaging measured residence time, spreading diameter, and rebound dynamics, while bottom-view imaging quantified the contact area. Infrared thermometry provided spatial heat flux distribution and overall heat transfer effectiveness. With increasing wall temperatures, distinct transitions between nucleate boiling, bubbly boiling, oscillating boiling, fingering boiling, and film boiling were identified. The combined visualizations provided detailed insights into effectiveness variations across boiling regimes, improving the understanding of droplet-wall heat transfer mechanisms. These findings support enhanced PCT modeling, advancing nuclear reactor safety analysis.

The present study addresses the critical need for accurate void fraction predictions in the engineering design and safety assessment of space-related two-phase systems. It investigates the drift-flux correlation under microgravity conditions, ranging from bubbly to annular flow regimes. This study compiles 458 experimental void fraction data points, revealing that distribution parameters vary with flow conditions and that drift velocities are minimal under microgravity conditions. Existing drift-flux correlations are found inadequate for capturing these variations and lack a simple model for drift velocity in microgravity two-phase flow. To address these issues, a new drift-flux correlation is proposed, considering flow condition effects on asymptotic distribution parameters and incorporating effective body acceleration to account for drift velocity decay in annular flow. The new correlation demonstrates strong predictive capabilities when evaluated against the collected experimental data, offering a significant advancement for space applications.

U-bend geometries are commonly used flow restrictions in nuclear reactor systems. Two-phase flows through U-bend are quite different from those in straight pipes. However, there has been no systematic study about inverted U-bend effects on two-phase flows. In the present study, a new experimental database is established using the existing Purdue University separate-effects test facility, featuring a 25.4 mm inner diameter pipe and a U-bend with curvature to diameter ratio R_c/D of 9. Detailed data including void fraction, gas velocity and bubble diameter are measured with miniaturized four-sensor conductivity probes with pressure loss obtained using pressure transducers. Using the obtained experimental data, mechanistic models have been developed to characterize the U-bend effects, which include models for pressure loss, U-bend dissipation length, variance of void fraction σ² and bubble velocity. It is found that the Lockhart-Martinelli’s two-phase flow frictional loss correlation can be used to predict the experimental two-phase pressure drop across U-bend with some modifications. The U-bend strength can be represented by the variance of the void fraction which dissipates exponentially in the U-bend dissipation region. The dissipation lengths of U-bend effects under different test conditions are determined by the dissipation rate β. The bubble velocity models are related to the development of σ². A modified Froude number Fr_m derived from the two-fluid model momentum equation is used as a fundamental parameter in developing the modeling correlations for σ², β, U-bend dissipation length and bubble velocity. All the modeled parameters can generally be predicted within an accuracy of ±10%.

In Boiling Water Reactors (BWRs), steam–water annular flow occurs near the fuel rods and plays a significant role in the nuclear reactor safety since the dryout of the liquid film may lead to the burn out of the fuel rods. However, the direct visualization and detailed liquid film measurement of high-temperature and high-pressure steam–water annular flow have been highly challenging due to the extreme operating conditions of BWRs (285°C and 7 MPa). This study addresses this limitation by developing a novel HFC134a–ethanol annular flow system at lower temperature and pressure (40°C and 0.7 MPa), effectively simulating the steam–water annular flow under BWR conditions. The experiments of HFC134a–ethanol upward annular flow were conducted in a 5 mm inner diameter tube using the constant electric current method and high-speed camera to obtain the liquid film thickness and flow behavior. The flow characteristics including base, average, and maximum film thickness and height of disturbance waves were obtained. Previous predictive models for these flow characteristics were tested with our measurement results. Through this simulating method, we report flow behaviors in detail achieving significant insights into liquid film behaviors in the actual BWRs.

U-bends are commonly used as flow restrictions in nuclear reactor systems, where two-phase flows exhibit significantly different behavior compared to straight pipes. Despite their importance, the effects of U-bends on two-phase flows remain underexplored. Interfacial area concentration (a_i), a fundamental parameter in two-fluid models, govern the interfacial transfer. The interfacial area transport equation (IATE) provides a superior approach to modeling a_i changes compared to conventional flow regime-dependent methods. In this study, IATE closure models have been developed using experimental database from the Purdue University separate-effects test facility, which features a 25.4 mm inner diameter pipe and a U-bend with curvature to diameter ratio R_c/D of 9. Experimental data suggests strong correlation between the variance of void fraction and covariance of Random Collision (COV_RC) in the U-bend and U-bend dissipation region. A modified Froude number Fr_m is used to model COV_RC. While constant values of covariance of Turbulent Impact (COV_TI) are used based on experimental results. Models for bubble velocity and pressure loss can be found in a separate U-bend geometric effects study. Void fractions are then determined using the continuity equation. Conventional drift-flux models are used in the straight pipe sections. Model coefficients of different bubble interaction terms are determined by evaluating each region (i.e., vertical upward, U-bend, U-bend dissipation, vertical downward) using experimental data individually. The one-group interfacial area transport along the whole test section is then evaluated using all the above closure models. The evaluation shows that the models predict a_i development accurately, with deviations generally within ±15%.

This work presents a mechanistic model for estimating the local heat transfer coefficient (HTC) in annular two-phase flow. The model is derived using fundamental principles and validated using data from two experimental facilities with different flow configurations and working flu-ids. Liquid-film thickness measurements were conducted using refrigerant at the University of Wisconsin-Madison, while HTC measurements were taken at an MIT facility using water as the working fluid. Non-invasive techniques are used at both laboratories to ensure the flow field is not disturbed. The physics-based modeling performed in this work ensures heat transfer performance in annular flow applications can be predicted with confidence.

The pool boiling experiments were conducted on a surface with micro-cavities, where the pitch and depth of the micro-cavities were kept constant at 120 μm and 20 μm, respectively. The cavity diameter (CD) ranged from 5 to 70 μm, and visualization was performed using visible ray and X-ray. The results of the experiments showed that at low heat flux, the heat transfer coefficient (HTC) was highest in the CD10-20 range, and there was a proportional relationship between nucleation site density and HTC. At high heat flux, excluding CD5, there was a trend of decreasing HTC with increasing cavity diameter. In this case, there was no significant difference in bubble behavior across the entire surface, and it is speculated that HTC decreases as the area of cavities filled with bubbles increases. Additionally, CD5 exhibited different bubble behavior compared to surfaces with CD10 and above, requiring different interpretation.

Capillary wicking can transport water effectively and can be applied in thermal management. Nanostructured materials have strong capillary wickability, which can promptly furnish water to the heating surface and enhance its boiling heat transfer characteristics. However, when the nanostructured surface is placed in an atmosphere environment, it will continuously adsorb volatile organic compound (VOC) from air, which can lead to a wicking deterioration. Therefore, this article mainly explores the effect of VOC adsorption or removal on the capillary wicking and boiling heat transfer characteristics of nanostructured materials. The contact angle and XPS results indicated that VOC adsorption could increase the pollutant content and hydrophobicity of nanostructured surface, while argon plasma treatment could remove VOC and enhance hydrophilicity. The pool boiling experiment showed that capillary wicking can greatly improve the threshold of boiling heat transfer and critical heat flux (CHF), while VOC adsorption can lead to a decrease in the capillary wicking performance of nanostructure, which can cause a deterioration of boiling heat transfer characteristics. After removing some VOCs through argon plasma treatment, the wicking and boiling heat transfer characteristics of nanostructure were partially restored. This work is beneficial for promoting the understanding of the impact of VOC on the heat transfer characteristics of wicking structures.

As a type of laser modification method, femtosecond laser surface modified technology can fabricate microstructures on surface of stainless steel, zironium alloys and nickel alloys, with special surface morphology such as honeycombs, humps and grooves. Modified surface with microstructures has significant impact on the heat and mass transfer. In order to explore the mechanism of enhanced boiling heat transfer on modified surfaces, visual research was conducted on nucleated bubbles on modified surfaces. It is possible to obtain the unique bubble behavior of modified surfaces, reveal the mechanism of enhanced heat transfer, and construct a model for bubble nucleation.Research can provide technical guidance and data support for the application of femtosecond laser modification technology in typical channels and various heat exchange devices.

Surface engineering has demonstrated significant potential for enhancing nucleate boiling heat transfer performance. However, the underlying mechanism remains unclear, especially the role of microlayer evaporation underneath bubbles. In this work, we systematically investigate the effect of surface micro-pillars on the microlayer morphology and the corresponding microlayer heat transfer performance. Using Direct Numerical Simulations, the microlayer formation and evaporation in the early diffusion-controlled bubble growth stage on various micro-pillar arrayed surfaces are reproduced. We reveal three distinctive microlayer morphologies on the micro-pillar arrayed surface: the disturbed microlayer, disrupted microlayer, and undisturbed microlayer. In general, a disrupted microlayer results in a reduced average thickness, increasing the transient heat transfer coefficient. Conversely, a disturbed microlayer retains more liquid, enhancing the microlayer heat transfer potential throughout its life cycle. Isolated bubble nucleate boiling experiments are performed to examine and further extend these findings throughout the entire bubble life cycle in nucleate boiling. The bubble dynamics are statistically analyzed. In addition, a preliminary experiment using synchrotron X-ray imaging is performed to directly capture the microlayer morphology. The experimental results align closely with the simulation results. Moreover, the experimental results indicate that a critical average microlayer thickness can be achieved by optimizing surface modifications, ensuring efficient evaporation throughout the bubble life cycle without significant depletion. This work provides a novel and practical way to optimize surface engineering for enhanced nucleate boiling heat transfer.

Heterogenous nucleate boiling observed in industrial scenarios, such as pressurised water reactors, originates at the smallest temporal and spatial scales. On the microscale, the evolution of the liquid-vapour interactions are directly influenced by the morphology of a heater’s surface. At this scale, nucleation mechanisms such as gas/liquid trapping, interface retention, and subsequent nucleation site activation are driven by hydrodynamic and thermal effects. In particular, the hydrodynamic phenomena become more pronounced when examining surfaces with increased surface intricacies. For example, the porosity or roughness can result in dominating capillary forces with complex surface tension and vapour diffusion effects. This process is further complicated by related thermal effects such as Conjugate Heat Transfer (CHT) from the solid to the fluid phases. This work aims to quantify the effect of nucleation mechanisms at the microscale and their impact on thermal efficiency and subsequent nucleate boiling. The progression of nucleate boiling is examined, using detailed Computational Fluid Dynamics (CFD) and the Volume of Fluid (VOF) method. Low capillary number simulations are performed using surfaces with porosity and roughness representative of the zirconium oxide layer on the cladding found in water-cooled nuclear reactors to discern more about the hydrodynamic and thermal physical processes that occur on such a surface. This is of particular interest, due to the challenges associated performing prototypic measurements of a water-cooled nuclear reactor across the range of length- and time-scales.

Point of Net Vaper Generation (PNVG) marks the transition in the bulk flow from single-phase to two-phase. Its prediction is paramount to modeling subcooled boiling and has therefore drawn consistent research attention. By searching for the incipience of bubble detachment via high-speed imaging, a recent flow-boiling dataset has identified hydrodynamically controlled PNVG in an upward annular channel. These data - which feature near-inlet PNVGs and a partially heated wetted perimeter - are compared with predictions by several well-known correlations. Some tested correlations were developed based on the dedicated mechanism of bubble detachment, and most of them claimed acceptable performance in fitting their own benchmarks. However, noticeable discrepancies are often found between the current data and these correlations, as well as among different correlations. In addition, certain correlations were originally reported with acceptable uncertainties in terms of dimensionless groups. These uncertainties are found propagating to unsatisfactorily low confidence in more sensible dimensional parameters such as the PNVG location, due to the nature of current conditions. These discrepancies and challenges faced by existing correlations are discussed with quantification and reasoning. Improvement is also attempted, and an analytically modified Levy’s model achieves acceptable consistency with the current experiment. The limited capability in predicting PNVG in the current configuration needs awareness and calls for further modeling improvement and validation.

During two-phase annular flow boiling at prototypical conditions in nuclear reactor fuels, liquid transport to the regions prone to CHF is an important consideration in the mechanistic prediction of dryout. Water flows to the CHF-prone region through a base liquid film flowing along the fuel sheath and through droplets that are entrained in the vapour flow, while water can be removed from the base film through entrainment processes. The base-film flow rate is related to the liquid superficial velocity while the droplet velocities move near the vapour core velocity. Recent work has shown that the presence of large liquid interfacial waves may be responsible for a large fraction of the liquid transport to the CHF region and with a speed in between that of the liquid and vapour. This paper presents the results of a development program to measure these interfacial phenomena in steam water boiling flows in a heated tube. A high output X-Ray source and single-photon detector counting array are used to record the characteristics of the interior flow patter at a speed of 250 frames per second. X-Ray energy was optimized to ensure transmission through the Inconel test section wall as well as to ensure differentiation in the cross sections of liquid and steam. Flow rates were limited to those where interfacial wave speed was less than 4m/s due to frame speed limitations. Measurements include wave speed, heigh, mass flow, and frequency. Important observations on the relationship between wave height, speed, and frequency will be presented.

Passive systems (PSs), both for reactor driving system and safety features, are currently significantly discussed because they will surely take an important role in the coming reactor designs including small modular reactors (SMRs), irrespective of water-cooled or non-water-cooled ones. Great many kinds of developmental effort are underway including the confirmation of their performances especially in the reactor safety aspects. While PSs should have many favorable points, their stability and thus reliability have been discussed with some concerns because of their small driving force especially in the coolant injection capability to maintain integrity of nuclear fuels under any kinds of accidents. This paper revisits two past efforts: ROSA-AP600 and passive containment cooling system (PCCS) horizontal heat exchanger for advanced boiling water reactor (ABWR) containment vessel under severe accident conditions, thus reactor system response and component performances of PSs, to consider key thermal-hydraulic capabilities assessed to assure the reliability in such a practical use of PSs.

Helical Cruciform Fuel (HCF) has a cruciform shape with helically twisted surface, which provides about 35% larger heat transfer area compared to standard cylindrical fuel. From a thermal-hydraulic perspective, this geometry results in a lower wall average heat flux leading to a power uprate potential. This study investigates two key thermal hydraulic-related phenomena for the HCF rods using Computational Fluid Dynamics (CFD) simulation: Departure from Nucleate Boiling (DNB) and fluid-elastic instability. The NuScale-like Small Modular Reactor (SMR) with a low mass low rate is considered as a reference plant design.

First, a numerical boiling test is conducted for a hot fuel pin using a Eulerian-based two-fluid approach, using the CASL boiling models to estimate the Critical Heat Flux (CHF) and Minimum Departure from Nucleate Boiling Ratio (MDNBR). Additionally, post-CHF cladding surface temperature is estimated to provide boundary conditions for a future fuel performance analyses. The performance of HCF is compared with that of standard cylindrical fuel under the same conditions to assess its relative advantages.

Furthermore, a Fluid-Structure Interaction (FSI) simulation is performed to estimate the Fluid-elastic Instability Margin (FIM) in the HCF geometry through vibration analysis. An unsteady simulation is carried out for a 2x2 lattice using the STRUCT-𝜀 turbulence model, which can capture vibrations without the need for LES-level mesh refinement. By using the vibration frequency and damping ratio—computed using the displacement data from the simulation —the FIM and fretting wear rate are estimated.

A global research effort is underway to develop Small Modular Reactors (SMRs) with diverse applications beyond just providing baseload power, while simultaneously enhancing safety. One design of interest is the NuScale concept reactor design, which utilizes natural circulation driven by temperature differences in the primary system, allowing for operation without pumps. This design has attention in worldwide for its high safety profile. However, a significant limitation of the natural circulation reactor concept is the lack of extensive operational experience. To overcome limitations, it is crucial to construct and operating scaled experimental facilities that can simulate natural circulation. In this study, the URI-SMR (UNIST Reactor Innovation-SMR), a scaled-down experimental facility based on the NuScale concept design, was employed to evaluate the natural circulation performance. The URI-SMR is well-suited for natural circulation study because its primary system is constructed of acrylic, which enables simultaneous performance evaluation and visual observation of the natural circulation flow. Through the URI-SMR, steady state experiments at various power levels were conducted, and the feasibility of load following operation being considered for SMR was also evaluated. In addition, the comparison between experimental results and system code analyses enhanced the reliability of system code modeling and established a foundation for the analysis of transient behaviors that are challenging to try in experiment. This research validates the natural circulation operational performance of integrated SMR designs like NuScale concept and extend confidence in next-generation SMR options, such as load-following capabilities.

Based on large PWR and SMART SMR (Small Modular Reactor) technologies, the innovative SMR,referred to as the i-SMR, is under development. The i-SMR incorporates an in-vessel helical coil steam generator.
Boiling in helical tubes differs from straight tubes due to the interplay of centrifugal and buoyancy
forces, which influence pressure drop and flow regimes. Heat transfer models used in safety analysis codes are largely based on straight-tube boiling phenomena, offering limited applicability to helical tube heat transfer predictions. Furthermore, flow instabilities under extreme conditions may compromise the structural integrity of pipes due to vibration, affecting flow stability, flow rate, and heat transfer performance, thereby potentially impacting system safety. To optimize performance and validate heat transfer in helical coil steam generators under conditions similar to the i-SMR, heat transfer tests are being conducted, and a test database is being developed. Additionally, to ensure the safety of integral reactor designs like the i-SMR, ongoing assessments aim to refine and evaluate heat transfer and pressure drop models specific to helical coil steam generators, which significantly differ in geometry and heat transfer characteristics compared to large-scale nuclear steam generators.
This study serves as an initial step in validating the heat transfer performance of the helical coil steam generators for the i-SMR. It focuses on evaluating existing heat transfer models, based on prior research, to determine their applicability to helical coil steam generators.

The passive auxiliary feedwater system (PAFS) is one of the advanced safety systems in the innovative Small Modular Reactor (i-SMR). The PAFS has a heat exchanger tube bundle submerged in the emergency cooling tank (ECT). In the PAFS, the heat is removed by the condensation in the heat exchanger tube having 3 degree inclination. To know the heat removal performance of the PAFS, the heat transfer rate for the condensation in tube should be accurately predicted. In this study, to evaluate the predictability of SPACE code for the condensation in the PAFS, the SPACE code analyses were conducted for the PASCAL and PICON experiments. For the PASCAL experiments which simulated a PAFS heat exchanger tube and the PICON experiments which simulated the condensation in the nearly horizontal tube, the heat transfer coefficient and flow regime were compared between the experimental results and the SPACE code analysis results. The SPACE code predicted well the condensation heat transfer rate in the PASCAL and PICON experiments when the condensation model developed by Ahn et al. (2014) was applied.

Supercritical carbon dioxide (sCO₂), nitrogen (N₂), and helium (He) Brayton cycles are promising power conversion systems for advanced nuclear reactors, including molten salt reactors (MSRs), sodium-cooled fast reactors (SFRs), and gas-cooled reactors (GCRs). Recuperators play a crucial role in enhancing thermal efficiency in these cycles but require volume minimization due to their high thermal duty and large size. This study investigates the thermal-hydraulic performance of straight-channel recuperators for sCO₂, N₂, and He Brayton cycles. Gas-to-gas experiments were conducted using printed circuit heat exchangers (PCHEs), covering a wide range of Reynolds (Re) numbers from laminar to turbulent regimes to accommodate various design conditions. Experimental results were analyzed based on Re numbers, and existing thermal-hydraulic correlations were evaluated for their applicability in recuperator design. In laminar regime, the developing flow effects are important for heat transfer and pressure drops. In transition and turbulent regimes, existing correlations have enough predicting performances. With the evaluated correlations, a validated one-dimensional (1-D) in-house PCHE design code was employed to determine the optimal recuperator volume while satisfying target effectiveness and pressure drop constraints. The optimal design results, derived under fixed thermal duty and pressure drop conditions, were examined across different Brayton cycle working fluids. The findings provide insights into the thermal-hydraulic performance of straight PCHE channels across a broad Re number range and offer valuable design-level guidance for recuperators in gas Brayton cycles. It is worth noting that these results contribute to improving the efficiency and feasibility of compact recuperators for advanced nuclear power systems.

Post-Critical Heat Flux (Post-CHF) is one of the most complex two-phase phenomena significantly affecting the coolability of nuclear fuel during a loss of coolant accident (LOCA) in light water reactors. Wall heat transfer characteristics in inverted annular film boiling (IAFB), inverted slug film boiling (ISFB), and dispersed flow film boiling (DFFB) regimes have been widely investigated in the literature to develop heat transfer models/correlations to predict the peak cladding temperature among other parameters. However, lack of comprehensive data necessary for validating physical assumptions made during modeling of the IAFB/ISFB/DFFB regimes has led to limited predictive capabilities of existing models and correlations. In this study, a series of quasi steady-state IAFB, ISFB, and DFFB experiments were performed in the Post-CHF Heat Transfer (PCHT) test facility at the University of Michigan that employs a direct hot-patch technique to stabilize the quench fronts in a tubular test section made of Incoloy 800H with an inner diameter of 12.95 mm. Experimental conditions spanned over a relatively broad range to investigate the effects of the liquid subcooling, mass flux, and system pressure on heat transfer in those regimes. Detailed test section wall temperature was acquired using thermocouples and the void fraction in the test section was measured using a gamma densitometer and an X-ray radiography system. In addition, the predictive capabilities and limitations of the existing models for those regimes were evaluated using the acquired film boiling experimental data. Based on the model benchmark results, improvements were proposed to enhance the model accuracy for IAFB/ISFB/DFFB wall heat transfer.

[1] This abstract is intended for Dr. Sun’s American Nuclear Society Thermal Hydraulics Division Technical Achievement Award (ANS THD TAA) Lecture.

This abstract was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product, or process disclosed in this paper, or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the U.S. Nuclear Regulatory Commission.

The unexplored potential of compact heat exchangers for use in light water small modular reactors offers a promising area for improving nuclear technology. Micro-channel heat exchangers provide high thermal efficiency and compact designs, making them suitable for integral designs. However, there is a significant gap in understanding their performance under liquid-boiling conditions, and no comprehensive database currently exists. This highlights the need for more research.

This study focuses on a pre-test analysis of a microchannel heat exchanger from the E-SMR database, developed within the ELSMOR project for a light water small modular reactor. Two models are used to simulate the performance of the heat exchanger: one with RELAP5 and the other with computational fluid dynamics (CFD) using STAR-CCM+. The comparison between the two codes addresses the limitations of thermal-hydraulic system codes like RELAP5 in accurately modeling microchannel heat exchangers, prompting the need for CFD to improve confidence in the simulations.
The pre-test analysis shows good agreement between the models. These will be validated against experimental data from a facility being built at Politecnico Milano to test compact heat exchangers.

The coupling of neutronic and thermal-hydraulic phenomena is important for the multiphysics modelling of the transient behaviour of nuclear reactors (e.g., design basis accidents). In this study, we analyse the behaviour of soluble-boron-free (SBF) water-cooled small modular reactors (SMRs) using coupled neutronic and thermal-hydraulic models. These models are developed using the Serpent Monte Carlo neutron transport and the Sub-channel CFD (SubChCFD) coarse-mesh CFD (computational fluid dynamics) codes. Typical industrial nuclear thermal performance software utilise nodal neutron kinetics and sub-channel nuclear thermal-hydraulic methods to simulate transient behaviour, but these methods oversimplify the geometry and 3D behaviour of coolant flow.

Although CFD models offer a high-fidelity alternative, it is computationally demanding to perform reactor transients. This is due to the fine computational meshes required and the modelling of complex turbulent and multiphase flows. Recently, coarse-mesh computational fluid dynamics (CM-CFD) models have been developed to mitigate this issue. These models utilise a sub-channel-based “filtering” mesh upon which empirical frictional and heat transfer thermal-hydraulic correlations are computed. In addition, the CM-CFD models also solve the Reynolds-Averaged Navier-Stokes (RANS) equations on a coarse computational mesh. However, unlike other CM-CFD approaches, SubChCFD also integrates and utilises experimental data.

This paper uses the coupled model to perform a full-core steady-state simulation of an SBF small modular reactor nuclear fuel assembly design developed by Alzaben et al.

In a pressurized water reactor (PWR), unwanted boron dilution transients can be caused by a large safety or regulating valve opening. When diluted coolant flows into the reactor pressure vessel, the cross-flow of two coolants of unequal boron concentration locally decreases in the core, possibly inducing a prompt change of reactor reactivity with a high impact on safety.

Concerns have focused on the behavior of pressurized water reactors (PWR) operating with soluble boron fluid in the reactor coolant. For the current analysis, a comprehensive understanding of factors influencing flow mixing patterns and boron diffusion in the reactor core is pursued through numerical investigations utilizing 3D computational fluid dynamics (CFD) simulations. These simulations play a crucial role in enhancing reactor operation safety.

A scaled PWR reactor pressure vessel, simulating a one-loop with different initial conditions of flow rates and Reynolds numbers for both the coolant and the safety injection flow into the cold leg. The simulation utilizes a customized solver implemented in OpenFOAM that considers the boron transport model using a transient flow algorithm coupled with a standard k-epsilon turbulence model. Given the magnitude of these simulations. The results provide a 3D mixing pattern under boron dilution transient and agree with experimental data regarding the optimal conditions for the best mixing and diffusion behavior of boron distribution entering the reactor core. This occurs at a specific ratio of injection to Reynolds number in the cold leg.

Small Modular Reactors (SMRs) are gaining importance in addressing energy challenges. To facilitate the wider adoption of these energy systems, it is essential to develop simulation tools that accurately represent the SMR phenomena. In this context, multi-physics and multi-scale analyses provide deeper understanding of SMR behavior under accident conditions. In this study, the US-NRC neutronic core simulator code PARCS, the KIT in-house thermal-hydraulic code TWOPORFLOW, and the US-NRC system thermal-hydraulic code TRACE were utilized. The TRACE/PARCS/TWOPORFLOW coupling code was developed following the ICoCo methodology, which involves exchanging data fields through mesh interpolation. An explicit temporal coupling is implemented, on one hand PARCS and TPF solve the reactor core using a domain-overlapping approach. On the other hand, TRACE solves the rest of the primary circuit and the selected auxiliary systems using a domain-decomposition approach. The NuScale plant has been analyzed using this multi-scale, multi-physics tool, showing good agreement with the reference data. Future work may explore a semi-implicit temporal coupling to enhance simulation stability.

Small modular reactors (SMR) have emerged as next-generation nuclear power systems, offering enhanced safety, efficiency, and economic advantages. Among their critical components, helical coil steam generators (HCSG) have been extensively studied for their effective heat exchange capabilities. However, primary-side cross flow within HCSG could induce vortices and turbulent structures between tubes, resulting in non-uniform heat transfer and flow instabilities that negatively impact overall system stability. Large eddy simulation (LES) based computational fluid dynamics (CFD) can accurately capture this complex behavior but requires fine meshes and short time steps, leading to high computational costs. In this study, a deep learning-based reduced-order modeling (ROM) strategy is proposed to maintain both accuracy and computational efficiency in analyzing local flow regions between HCSG tube layers. Proper Orthogonal Decomposition (POD), Dynamic Mode Decomposition (DMD), and a nonlinear autoencoder are employed to reduce data dimensionality, followed by a Long Short-Term Memory (LSTM) network for predicting flow evolution. These ROM frameworks (POD-LSTM, DMD-LSTM, and Autoencoder-LSTM) are compared to identify the most effective approach for significantly reducing simulation overhead while preserving CFD-level predictive accuracy. The results indicate that linear methods effectively capture dominant features such as large-scale vortex formation and dissipation, whereas the nonlinear autoencoder emphasizes random flow diffusion and chaotic behavior. Notably, the POD-LSTM model demonstrates superior performance in predicting flow field dynamics, achieving higher coefficients of determination (R^2) compared to the other models.

The LUT Heating Experimental Reactor (LUTHER) is a Small Modular Reactor (SMR) concept designed for safe district heating. The project has a heat capacity of 24 MWth, which is sufficient for heating small communities and businesses. To improve the efficiency of heat distribution, it is crucial to locate the plant in close proximity to consumers. As a result, a primary design criterion is to maintain the highest standards of safety.

Many severe accidents in nuclear reactors, such as rapid reactivity injection and core meltdown, are largely prevented by the reactor core's design. To transfer decay heat to the ultimate heat sink after potential accidents, a fully passive system has been developed to transfer heat from the core to the environment through boiling, free convection, condensation and wall conduction, using bedrock as an intermediate heat sink. The presence of non-condensable gases significantly influences heat transfer and steam condensation, making the calculations more complex and design more challenging.

In this paper, heat fluxes in the heat exchanger from containment to bedrock were calculated and visualized using the TRACE version 5 system code software. The effectiveness of employing bedrock as a heat sink was evaluated, and essential design parameters for the heat exchanger were established. These parameters include optimal pipe spacing, the appropriate pipe depth for maintaining a low surface temperature, the pipe length and inclination angle to facilitate efficient condensate flow.

Rod bundle spacer grids are known to have important, possibly dominant effects on thermal-hydraulic behavior during core uncovery. Spacer grids enhance mixing and convective heat transfer downstream of the grid Droplet breakup at a spacer grid increases interfacial area and the de-superheats steam benefiting heat removal from uncovered rods. Because they are unpowered the spacer grids can also rewet much easier than surrounding fuel rods. The liquid film on a wet grid acts as a source for droplet entrainment immediately downstream of the grid which further increases interfacial area and steam de-superheat.

The Rod Bundle Heat Transfer Facility (RBHT) data provide a unique opportunity to determine conditions for spacer grid rewet during reflood. Each spacer grid in RBHT has one or more wall-mounter thermocouples that indicate if and when the grid rewets. Tests ranged in RBHT from low flooding rates (0.5 cm/sec) to 15 cm/sec. Previous studies have shown that spacer grids can rewet well ahead of the quench front on the heater rods. However, no systematic study has been done to characterize when and which spacer grids rewet. The effects of wetted grids on the droplet dynamics and two-phase heat transfer downstream of the grids have not been seriously investigated.

In this paper results from several RBHT tests are presented and discussed with an emphasis on spacer grid rewet and the challenges in simulating the rewet process and effects on transient reflood behavior.

Interface Capturing simulations are becoming a more practical tool for complex flow analysis due to significant improvement of flow solvers, pre- and post-processing tools as well as rapid development of high-performance computing capabilities. This creates exciting opportunities to study complex reactor thermal hydraulic phenomena. This presentation will focus on the history and review of numerical flow simulation approaches in recent years, capabilities development and validation as well as the applications to practical problems of interest. We will discuss typical computational methods used for those simulations, provide some examples of past work, as well as computational cost estimates and affordability of such simulations for research and industrial applications. New generation methodologies are required to take full advantage of those capabilities to greatly enhance the scientific understanding of complex flow phenomena in various conditions relevant to nuclear energy applications.

The core interfacial friction model plays a dominant role in the prediction of core cooling and maximum clad temperature in PWR and BWR accident sequences. Due to a lack of precise void fraction data in such complex geometry, the code models of core interfacial friction still have a rather high uncertainty band. The FONESYS network of system code developers initiated an activity to compare the models of the ATHLET, CATHARE, LOCUST, RELAP5 and SPACE system codes. A first comparison of models is made in the domain of low flowrate in conditions where wall friction is small and void fraction depends only on a balance between buoyancy force and interfacial friction. Only the pre-CHF bubbly-slug-churn and annular flow regimes without drop entrainment are considered in the domain of void fraction (0 < α < 0.8). The effects of pressure (0.1 MPa < P < 12 MPa) and hydraulic diameter are investigated. The paper presents first the origin of the models. The results show that the various codes agree on the qualitative pressure effects with some differences on the hydraulic diameter effect.
However, significant differences are found in the very low-pressure domain. Possible effects of the wall heat flux are discussed. Further work is planned including validation and possible model improvements.

The LWR thermalhydraulic behavior in accidental transients encounters many types of bifurcating events (BE) with cliff-edge effects where some phenomena disappear and other appear. They are due to automatic control or operator actions such as SCRAM, ECCS actuation, pump start or stop, valve opening, etc, or to transitions between different flow regimes or heat transfer regimes. Such BE have a major impact on the main parameters of interest such as primary (and secondary) pressure and fluid mass inventory, and on the figures of merit such as a peak clad temperature. The prediction of all BEs and of the right occurrence time of BEs is the main challenge of experimental and numerical simulation tools. In the Phenomena Identification and Ranking Table (PIRT), the successive time phases of a transient or Phenomenological Windows (PhW) are first identified and they can be defined and bounded by some BE. In the scaling analysis performed to design Integral Tests Facilities (ITFs) and Integral Effect Tests (IETs), most scaling methods use dimensional analysis of scaling equations (mass, momentum, energy) at system scale with acceptance criteria on the ratio of non-dimensional numbers at reactor and experimental scales. This forgets the dominant role of BEs. One may use a mature system code to perform a more detailed scaling analysis of a transient and to focus on the respect of occurrence and timing of occurrence of major BEs as acceptance criteria for the design of an ITF and of IETs. Examples are given on some PWR LOCA analyses.

FONESYS is an international network of system code developers created in 2010 to share information on R&D, to benchmark codes, to discuss the Validation and Verification as well as the code scalability and uncertainty quantification. APROS, ARIANT, ATHLET, CATHARE, CATHENA, COSINE, MARS, LOCUST, MARS-KS, RELAP5, RELAP5-3D, SPACE, TRACE are the codes that were involved in the FONESYS activities: updating the state of the art, identifying issues, discussing envisaged solutions, sharing experience in 3-field models, transport of interfacial area, numerical issues and well-posedness, code uncertainty evaluation. Code benchmarking were performed on boiling channel with CHF and Post-dryout, two-phase critical flow, flow regime transitions in horizontal flow, core interfacial friction, core 3D-mixing effects and two-phase singular pressure losses. Many code improvements were implemented in the various codes following the benchmark activities. When the need of new experimental data was identified, FONESYS discussed with SILENCE experimentalists to define requirements of new instrumentation and new experiments. Future activities will focus on code scalability, core 3D modelling and validation, two-phase pressure losses, use of system codes for scaling analysis and applications to passive systems, SMRs and AMR. The present paper presents the major achievements and the motivations for the future activities.

The system codes can model 3D flow in a core either with 3D solvers in porous medium, or with sub-channel models, or even using cross-flow junctions between parallel 1D models. Such tools rise many questions on the modelling of mass momentum and energy transfers including diffusion and dispersion processes. The extrapolation of many closure laws from 1D to 3D flow and the formulation of wall friction and interfacial friction when the flow is not parallel to fuels rods require some attention. The FONESYS network of system code developers initiated an activity to compare the 3D-core models of the ATHLET, CATHARE, COSINE, LOCUST, RELAP-3D and SPACE codes in very simple situations. One considers first two adjacent fuel rod assemblies with different power and one calculates the vapor flow above a swell level at two pressures (1 and 7 MPa) to obtain either friction-driven crossflows or buoyancy driven crossflows (chimney effect). Other calculations include the two-phase region, a swell level and a dry zone above, the assemblies being connected to an upper plenum and a lower plenum, allowing different inlet flowrates as in a reactor situation. The impact of axial friction pressure losses and spacer-grid form losses on radial crossflows is shown. The homogenization of void fraction below a swell level seems very efficient. The sensitivity on radial pressure losses in non-axial flow is not very high but the uncertainty on the radial wall friction and interfacial friction is very high. From the analysis of these results, further investigations are planned.

The current system codes use 2-fluid models or 3-field models and predict wall friction and singular pressure losses by models developed for a unique mixture momentum equation. Two-phase multipliers exist that can extrapolate 2-phase pressure losses from 1-phase models but the repartition between phases is not modeled although it may play a very significant role on the result. This may result in a rather high uncertainty of predictions not only in high velocity flow but also in low flow situations encountered in natural circulation. The FONESYS network of system code developers initiated an activity to compare the predictions in a few basic singular geometries for which one can simply evaluate the single-phase pressure loss coefficient. In a reactor circuit, there are several locations with an abrupt area restriction or an abrupt area enlargement or even some plates behaving as a diaphragm. Some code prediction comparisons in such basic geometries are presented and analyzed in both vertical upward and vertical downward flow and in five flow regimes: 1-phase liquid, two-phase bubbly flow, low velocity and medium velocity slug-churn flow and high velocity annular-mist flow. The calculations were made with five codes: ATHLET, CATHARE, MARS-KS, RELAP5 and RELAP5-3D. Some differences are found between codes. The effects of nodalization are investigated and the impact on void fraction perturbation are analyzed. First conclusions are drawn on the reliability of predictions and some requirements for future well-instrumented pressure loss experiments are defined.

The Atomic Energy and Alternative Energies Commission (CEA) is a French Research and Development institution that plays a major role in the French nuclear energy program. CEA has been developing the CATHARE code for 45 years. It is an extensively validated thermal-hydraulic system code based on the 2-fluid 6-equation model. The CEA is currently developing new tools in CATHARE to facilitate its use for scaling analysis, which contribute significantly in the identification of dominant phenomena occurring during reactor accidental transients and in the design of experiments able to simulate major phenomena with minimal distortions.

Previous published work on scaling analysis with the CATHARE code focused on the primary mass and primary pressure equations to identify the dominant terms controlling the mass inventory and the system pressure. These terms were calculated “by hand” during the post-processing phase. A new tool enables this analysis to be carried out automatically during a calculation, by fetching the thermal-hydraulics quantities at code execution. It relies on a Python supervisor, based on the ICoCo (Interface for Code Coupling) standard, which extracts the required fields at each calculation time step. The fields are then manipulated using the MEDCoupling open source library. The integrated momentum equation along cooling loops is also added to the analysis.

This work describes the way this tool is applied to a SB-LOCA transient using the equations of mass, momentum and pressure. Figures of merit and available post-processing are presented, which enables the dominant phenomena to be quantified. Leads for future developments are given.

The models in system codes for LWR core interfacial friction still have a rather high uncertainty in some conditions due to a lack of void fraction data in such complex geometry. Usually, only axial pressure differences are available to test the interfacial friction. The FONESYS network of system code developers initiated an activity to compare the models of several system codes and found significant differences at very low pressure. Recently a new void fraction measurement technique based on the wire mesh sensor was implemented by CRIEPI in a 5x5 rod bundle and produced real void data. There are steady state data at 1, 3 and 7.2 MPa, with a wall heat flux from 5 to 45 kW/m² and a mass flux in the range 90 to 400 kg/m²/s. CRIEPI also performed boil-off tests with an outlet atmospheric pressure and heat fluxes in the range 2.4 to 7.2 kW/m². EDF and CEA performed validations calculations with the CATHARE code on these data. Several sensitivity tests are presented to investigate some possible reasons of the code-to data differences. Previous validations on data in boil-up tests and on air-water non-heated rod bundles data are used to investigate the various possible effects of fluid properties, of pressure, and of the wall heat flux. Preliminary conclusions are drawn on the validity and limitations of the current interfacial friction model and on possible ways to improve it.

Advanced nuclear fuels—such as Accident-Tolerant Fuels (ATF), helical fuel, and Transformational Challenge Reactor (TCR) fuel—have underscored the need for high-fidelity multi-physics fuel performance analysis. To address this demand, the BEEs code (developed by the XJTU-NuTheL research group) offers a multi-physics, multi-dimensional simulation framework tailored for advanced nuclear fuel systems.

By integrating high-fidelity finite element models with advanced coupling strategies, BEEs tackles critical challenges in fuel performance evaluations under diverse operational scenarios. The code features comprehensive models for diverse fuel types, including UO₂-Zircaloy rod fuel, TRISO-coated particle fuel, annular fuel, and plate-type fuel, incorporating thermal-mechanical behavior, irradiation effects (creep, swelling, fission gas release). Validation studies demonstrate good agreement with experimental data and benchmark cases for predictions of temperature, stress, and deformation under normal operation, Loss-of-Coolant Accident (LOCA), and Reactivity-Initiated Accident (RIA) scenarios. Meanwhile, both discontinuous Galerkin and high-order implicit time-stepping methods have been introduced for one-dimensional coolant channel modeling with enhanced accuracy. Application highlights span fuel performance evaluations for ATF, annular fuel, gas-cooled reactor fuels, and plate-type fuel, alongside multi-physics coupled simulations of typical PWR primary circuits and plate-type fuel assemblies.

This work establishes BEEs as a promising code for serving as both a fuel design tool and an independent performance evaluation platform for solid rod fuel, annular fuel, plate fuel, and TRISO.

Small Modular Reactors (SMRs), particularly Light Water-SMRs, represent a viable option for near-term nuclear deployment in Europe, building upon established Light Water Reactor technology while incorporating evolutionary design modifications like passive safety systems. While these systems offer advantages such as independence from external power and component minimization, they face potential functional failures due to Natural Circulation issues. Therefore, passive system failures must be addressed in SMR design and safety reviews. Current guidance on passive safety system requirements and failure mode modeling methodologies needs consolidation. Short-term research priorities include reliability analysis of ThermalHydraulic phenomena driving system operation and related Uncertainty Analysis. Building on ENEA's MELCOR input-deck developed within the Horizon Euratom Safety Analysis of SMR with Passive Mitigation Strategies-Severe Accident (SASPAM-SA) project, this work applies the REPAS (Reliability Evaluation of Passive Safety Systems) methodology to the Emergency Heat Removal System (EHRS), a key passive safety feature for decay heat removal in advanced designs. REPAS will help analyze various system states, including low-probability scenarios, to understand EHRS behavior and its plant impact.

The flow characteristics in the upper plenum of molten salt (MSR) and high-temperature gas-cooled reactors (HTGR), during the pressurized conduction cooldown accident (PCC) scenario, are dominated by natural convection jets emitted from the top of the reactor core. Benchmark experiments are necessary to validate the Computational Fluid Dynamics (CFD) codes currently being used to further characterize this phenomenon. This study focuses on providing benchmark data for the upper plenum PCC scenario. The experimental facility is a scaled-down generic model of the upper plenum of MSRs and HTGRs. This study produces velocity field measurement data, via Particle Image Velocimetry (PIV). Data from optical fiber-distributed temperature sensors, thermocouples, and volumetric flow data are additionally used to calculate the conjugate heat transfer characteristics of the plenum and provide boundary conditions for the CFD models. The test matrix consists of isothermal and non-isothermal cases. In the non-isothermal case, fluid flow is driven by natural convection and buoyancy forces while the isothermal case is driven by pump-induced pressure gradients. This paper presents a detailed description of the experimental methods and analysis techniques utilized in this study and the results of multiple isothermal and non-isothermal test cases.

Deep pool reactors are reactors that operate at low pressures and are usually built near residential areas for heating and cooling, so the design of the reactor system requires high intrinsic safety characteristics. The design of non-energetic safety systems for deep pool reactors is characterized by multiple novel safety devices, and the transient operating characteristics of the devices are critical to the safety and stability of the reactors. In this paper, a set of experimental system that can reproduce the non-energetic waste heat export function of the pool reactor is set up, and the transient characteristics of the equipment under the loss-of-flow accident are experimentally investigated. The experimental results verify the on-time response of the relevant non-energetic equipment after the design accident. Extended working condition numerical studies were also carried out using a system analysis program.The experimental results demonstrate that the deep pool reactor can export the waste heat of the reactor through its own non-energetic safety system under a loss-of-flow accident.

Passive safety systems are an economically interesting alternative to conventional active systems, which are also more robust against many external influences, as they do not rely on an external, active drive. Accordingly, they continue to function even if large parts of the plant infrastructure are damaged or unavailable, as was the case in Fukushima Daiichi accident, for example. However, passive heat removal systems in particular pose major challenges for designers and the thermal-hydraulic calculation tools they use. One reason for this is the coupling or feedback between the heat flow that is introduced into the coolant and the mass flow of the coolant through the system, which results from the heat input. In addition, state of the art heat transfer models obviously cannot capture the heat transfer for passive systems precisely enough. As a result, attempts to recalculate experimentally determined heat transfer rates of passive heat removal systems with fluid dynamic codes have sometimes resulted in considerable deviations between experimental and numerical data. This paper presents two new heat transfer models developed specifically for passive systems. It is described how they can help to better calculate and predict the performance of related systems. The models were developed primarily on the basis of experimental data recorded at the COSMEA and NOKO test stands and published by the Helmholz Center Dresden Rossendorf. In addition, the model development was supported by CFD calculations executed to better understand the underlying mechanisms.

In the Small Innovative Helium-Xenon Cooled MObile Nuclear power System (SIMONS), the reactor core is horizontally oriented. To investigate the thermal-hydraulic effects of this configuration, a computational fluid dynamics (CFD) methodology is utilized. The flow and heat transfer dynamics of the core assembly under natural circulation conditions are examined, utilizing the passive gas circulation test loop at Shanghai Jiao Tong University as a reference model. The influence of heating power on mass flow rate, Reynolds number, pressure drop, and Nusselt number is examined, delineating the power range that enables flow self-compensation within the natural circulation system. Moreover, this investigation delved into the distribution patterns of mass flow rate, temperature and heat transfer across various channels under conditions of reduced natural circulation. The findings revealed that with a decrease in mass flow rate, there is a progressive increase in the proportion of flow rate and heat transfer within the lower-positioned channels, And the matching degree between the flow rate and heat exchange of each channel decreases. Furthermore, the study observed a pronounced thermal stratification in the upstream chambers at reduced flow rates, which can be attributed to the more pronounced heating exerted by the bottom of the assembly, coupled with the obstruction of flow in the upper chamber by the buoyant ascent of low-density Helium-Xenon mixture.

Small Modular Reactors (SMRs) rely on passive safety systems, which use gravity-driven natural circulation to transfer heat from the core to a large reservoir without operator intervention for extended periods of time. Some SMR designs can reach Rayleigh numbers around 10^15, but the limited experimental data at this scale highlights the need for further validation of predictive thermal hydraulic codes.

This study presents preliminary Computational Fluid Dynamics (CFD) calculations to support the design of an experimental rig at the PANDA facility at PSI, Switzerland, under the OECD/NEA PANDA project. Using a 2D axisymmetric CFD model created with EDF's open-source Code_Saturne software, parametric studies were conducted to investigate the heat transfer mechanisms. Key variables, such as geometrical configuration and dimensions, operating conditions, and numerical options, were examined. Initial results indicate that achieving a Rayleigh number of 10^15 is feasible under the constraints of the facility, with favorable alignment to existing Nusselt-Rayleigh experimental correlations at lower Rayleigh numbers. The computational results allowed the sizing of the components and fixing the operating conditions. These results also underscore the importance of a detailed temperature measurement strategy, particularly near the wall, to be used for the validation and verification of the models.

Further studies explore the influence of different turbulence models on the results, revealing notable differences in predicting the laminar-to-turbulent transition zone and the peak wall temperature. These findings are useful for ensuring the validity of Nusselt-Rayleigh correlations over a wider range of applicability and allowing for accurate modeling of large-scale passive safety systems.

The conventional containment vessel of nuclear power plant acts as a barrier against radiological impacts and missile threats, independent of the reactor’s cooling. However, as the size becomes more compact, the containment vessel of an SMR roles as the final passive cooling system. The cooling capability around the dome must be ensured during severe accidents, such as Loss of Coolant Accident, to minimize coolant loss through condensation on the dome. Most condensation occurs in the upper dome of the containment vessel, however, studies on the external cooling capability of the dome reflecting the large Ra of SMR have not been conducted. Therefore, this study aims to analyze the heat transfer of domes under high Ra conditions, considering their geometric characteristics and flow behavior. Experimental range corresponds to 10⁹≤Ra_Db≤10¹³. The base diameter of dome, corresponding to the diameter of the cylindrical lower structure, is fixed, but its height can vary. These characteristics of the dome can be defined by a truncation angle, as the dome represents a segment of a sphere. The smaller the truncation angle, the closer the dome resembles a flat plate, while a truncation angle of 90° represents a hemispherical shape. Depending on the truncation angle, the slope of the dome surface varies, leading to differences in flow behavior. Plume flow enhances heat transfer and the separation point, where the point plume flow develops, is varied by the surface slope. Smaller truncation angles result in an earlier development of the separation point and further enhance heat transfer.

Due to the nonlinearity of the Natural Circulation (NC) process, flow and temperature perturbations in an NC loop result in a feedback mechanism. Under certain conditions, the prevailing circulatory flow can become unstable and ultimately stall, leading to insufficient heat transfer. This scenario is of particular interest to PWRs, since inadequate passive heat removal could result in initiation of postulated accident scenarios. Therefore, accurate predictive capability for stall in perturbed NC scenarios is of great interest to this work.

Performing full-scale high-fidelity simulations of the plant during perturbed scenarios is computationally prohibitive from a design basis analysis perspective. However, the phenomena of interest during perturbed scenarios are highly complex and multiscale; current 1D System Codes are not capable of adequately capturing the prevailing 3D flow phenomena. The current work considers a more computationally feasible approach. We explore surrogate modelling via Gaussian Process Regression as a method for accurately predicting bulk flow and therefore NC stall. The surrogate model is trained by considering Unsteady Reynolds Averaged Navier-Stokes (URANS) Computational Fluid Dynamics (CFD) simulations of NC in a simple experimental loop geometry containing an n-bend and multiple heat sinks. Various perturbed NC simulations are presented, in which we consider multiple transient definitions which are specified by an injection flow rate profile at a specified location, heater input power, and n-bend height. We then assess the surrogate modelling approach’s ability to capture the key features of quantities of interest by benchmarking the results against equivalent URANS CFD simulations.

As the primary mechanism of passive heat removal in nuclear reactors, understanding natural circulation is paramount to their safe operation and shutdown conditions. Natural circulation in molten salt reactors is of particular relevance due to the fluids high Prandtl number as well as the phase of the fissile material in liquid-fueled reactors. Natural circulation loops are employed to study the thermal hydraulic behavior of fluids when subject to thermal gradients and small flow disturbances. The objective of this work is to introduce and analyze the effects of local heating conditions on the velocity profile and near wall behavior (such as boundary layer thickness) of molten salts in a heated transparent test section. Particle image velocimetry (PIV) was performed, and the boundary layer was analyzed for three different heating conditions. Those conditions were applied to the transparent test section: a cooling test condition, a thermally isolated test condition, and a heated test condition. In addition, the other thermal conditions of the loop were held constant. As the test section is heated, the peak velocity and slope of the velocity profile increase with test section heater power. Additionally, preliminary transient thermal analyses are presented in this work.

The passive molten salt fast reactor (PMFR), under development in Korea, employs a helium bubbling system to remove insoluble fission products (IFPs). Notably, the helium injection significantly changes entire fluidic performance within the primary system. These changes, in turn, influence heat transfer efficiency. Accordingly, a lab-scale experiment needs to be performed to evaluate the impact of helium injection under PMFR conditions.

To this end, a two-phase natural circulation molten salt loop at a reduced scale was designed based on scaling laws to simulate the helium bubbling effect in the PMFR. New governing equations for the prototype (PMFR) and reduced model (molten salt loop) were established based on the drift-flux model. Similarity criteria were derived through the nondimensionalization of the new governing equations. These criteria were employed in the design of the reduced model. Furthermore, an enlarged model, whose geometrical shape is similar to the reduced model, was also designed to evaluate the effect of friction number.

Subsequently, the distortion among prototype and two models was evaluated using multiphaseEulerFOAM solver in OpenFOAM. The distortion analysis revealed significant discrepancies in void fraction and fluid velocity between the prototype and the reduced model. A major reason for the distortion is attributed to geometrical differences. However, the distortion between the reduced model and the enlarged model was relatively minor due to geometric similarity and friction number scaling. This study will contribute to developing advanced scaling laws applicable to molten salt reactors (MSRs), although there are still rooms for improvement.

The steady-state behaviour of two-phase natural circulation systems (TPNCs) is critical for their efficient and safe operation across various applications, including nuclear reactors, thermal power plants, and advanced passive cooling systems. In TPNCs, steady-state conditions represent a balance between the buoyancy driving force and the opposing frictional force. Understanding and predicting the steady-state characteristics are vital for optimizing system performance, particularly in Boiling Water Reactors (BWRs) and natural circulation boilers (NCBs), where high flow rates are desirable for enhancing the heat transport capability.

This paper provides an in-depth analysis of the steady-state behaviour in TPNCs, focusing on key factors such as loop flow regimes, the effects of system geometry and heat input on circulation patterns. The study examines the influence of system parameters like loop diameter, heat flux and pressure on steady state flow. In addition, the effect of loop inventory on the steady state performance has been studied. The predictions cover the inventory at which peak TPNC flow rate, breakdown of TPNC flow and heat-up of the heated surface occurs. Insights gained from this analysis are crucial for designing TPNC systems to maximize the heat transport capability for critical applications such as nuclear reactors and thermal power plants.

This paper presents the development and validation of a high-fidelity thermal-hydraulic model of a molten salt natural circulation flow loop, designed for integration into a digital twin framework. This paper compares OpenFOAM and Idaho National Laboratory’s code Pronghorn against experimental data from Texas A&M University’s (TAMU) Molten Salt Flow Loop (MSFL).

Validation includes three natural circulation test cases: two-dimensional single-phase, two-dimensional with bubble injection, and three-dimensional single-phase flows. Key figures of merit include qualitative flow profile, accuracy of steady state temperature prediction, and computational efficiency for assessing the codes’ performance.

Preliminary OpenFOAM and Pronghorn results for two-dimensional single-phase agree qualitatively with flow profile. Properly accounting for the experiment’s unaccounted heat losses and the high computational cost have been the biggest obstacle to full validation. Additionally, Pronghorn’s PIMPLE algorithm is under rapid development, with heat-flux boundary conditions to be added in the near future. Initial three-dimensional single-phase models currently exhibit prohibitively-high runtimes and computational costs.

Before the final submission, we will improve model performance to adequately simulate the steady-state single phase models in two and three dimensions, as well as endeavor to implement the proper bubble boundaries. Future work will explore complexity reduction techniques for implementation of a fast-running model within a digital twin framework.

The modeling and simulation of Molten Salt Reactors (MSRs) require a comprehensive multiphysics approach to capture the complex interactions between thermal-hydraulics, neutronics, structural performance, and salt chemistry. This paper introduces an integrated multiphysics modeling framework to support MSRs development using Idaho National Laboratory (INL)’s Pronghorn.

At the core of this framework, thermal-hydraulics is coupled with neutronics, enabling accurate predictions of the dynamic behavior of the MSR core and fuel salt under varying operational conditions. The framework includes detailed neutron transport models combined with weakly compressible thermal-hydraulics models for fuel salt with void transport. Additionally, corrosion modeling, informed by thermochemistry, simulates material degradation and its long-term impact on reactor performance. Furthermore, the integration of redox potential control provides a crucial mechanism for regulating corrosion rates and maintaining fuel salt chemistry stability. This comprehensive approach facilitates the evaluation of safety margins, optimization of reactor designs, and development of strategies to minimize corrosion and ensure long-term reactor reliability.

This integrated approach is unique and novel and is being applied to the Molten Chloride Reactor Experiment (MCRE) design and analysis which demonstrates its practical utility. The results are significant for advancing the safety, performance, and sustainability of MSR technology, reinforcing its potential role in the future of clean energy production.

The Molten Salt Fast Reactor (MSFR) design has the particularity that the fuel is the coolant itself. This produces a tight coupling between neutronics and thermal-hydraulics as the fuel circulates through the primary system. Therefore, developing computational models for the analysis of the MSFR requires a multi-physics approach.

The fission process generates fission products, some of them which decay releasing both decay heat and delayed neutrons. These are known as delayed neutron precursors and decay heat precursors (DNPs), respectively. In the MSFR, these precursors originate and are carried by the liquid fuel throughout the primary circuit. The generation, transport, and decay of the DNPs affect the neutron flux, heat source, and temperature distributions in the MSFR.

In the research, we propose to develop a neutronic – thermal-hydraulics computational model of the MSFR that considers the transport of the delayed neutron and heat precursors along the primary circuit. The principal computational tool chosen for this purpose is the high-fidelity code Cardinal, a wrapping within the MOOSE framework that integrates the Computational Fluid Dynamics code NekRS and the Monte Carlo particle transport code OpenMC.

Flow distribution through the core of a nuclear reactor is a key consideration when predicting full field temperatures and pressure drops. While these metrics are also dependent on the neutron flux distribution in any reactor, a liquid fueled molten salt reactor presents added complexity due to the fact that the heat is generated in the flowing fluid itself. The Natura Resources’ MSR-1 design calls for support grids in the upper and lower plenum of the reactor, which in turn can significantly impact the flow distribution throughout the core. Numerical modeling is performed on a one quarter core with five different grid cases using ANSYS FLUENT with an inlet pipe Reynolds number equal to 1.7E4. The baseline case considers the geometry with no support grids. Each grid is represented as a radially weighted porous media with greater porosity at the extremities to facilitate uniform flow distribution. The deviation of the results from the baseline case are determined and a relationship for predicting a given channel’s mass flow as a function of the grid porosity is proposed.

The liquid-fuel molten salt reactor (MSR), uniquely employing molten salt as both nuclear fuel and coolant, exhibits distinct thermal-hydraulic characteristics due to internal heat generation during flow. This study investigates the flow and heat transfer behavior of molten salt with internal heat sources using CFD simulations. Results reveal significant deviations (up to 52%) in the Nusselt number predicted by traditional correlations (e.g., Gnielinski) for transition flow regimes (Re = 5×10³–1×10⁴), while the Di Marcello model reduces errors to 15%. Friction pressure drops align with classical models (Blasius, Guo and Julien，McAdams) within 17% deviation. In addition, microwave heating is proposed as a new internal heat source experimental method to verify the influence of heterogeneous power distribution on Nu number (less than 33% deviation). The results provide a basis for the thermal design and experimental method optimization of molten salt reactor.

Molten Salt Reactors (MSR) make for a promising technology for nuclear reactor design, due to their flexibility, inherent safety features and waste-burning capabilities. For unmoderated MSRs, the core consists of a large vessel without internal structure guiding the fluid, characterized by very high Reynolds numbers and a highly turbulent salt flow. Moreover, in those reactors, neutronics and thermal-hydraulics are strongly coupled physics due to the significant thermal feedback coefficients and need to be considered together. In previous studies on the Molten Salt Fast Reactor (MSFR) concept, the flow used to be computed with Reynolds Average Navier-Stokes models, which are unable to capture the temporal fluctutations. More recent studies applied a Detached Eddy Simulations (DES) calculation to address this problem and optimized the power stability. However, those studies were using wall functions and high aspect ratio cells in order to reduce computational cost. This led to low precision and prevented eddy computations in this region, resulting in an apparent viscous layer damping all temperature variations. Consequently the wall temperature remains an open question.
In this work, we aim to investigate the impact of resolving near wall region with LES models in specific regions to assess if the temperature fluctuations are potentially inducing thermal fatigue in the vessel of the MSFR. These studies are conducted with coupled neutronic-thermal-hydrauylics simulations on the full core of the MSFR, using Ansys Fluent for the thermal-hydraulics part and the Transient Fission Matrix approach for the neutronics.