Conference Agenda

The Nuclear Energy Agency (NEA) and its Committee on the Safety of Nuclear Installations (CSNI) has been pioneering since 1970 in advancing thermal hydraulics and reactor safety assessment through various projects and methodologies, significantly contributing to T/H nuclear safety research. Despite challenges, the progress made provides a strong foundation for future advancements in nuclear safety research and application.

A renewed effort is required to extend thermal hydraulics research beyond traditional water-cooled reactors (WCR) to non-WCR and advanced reactor designs and beyond the traditional end-uses of nuclear. The extensive experience gained from water-cooled nuclear reactors is fundamental in advancing the safety and reliability of next-generation nuclear technologies.

Experimental data are precious, rare, expensive and yet still needed! The first priority is preserving the competences, expertise and database derived from past research investments. And such data needs to be well known and used! It is also fundamental to create new ones in a prompt manner, because “the train was already leaving the station”. Defense in depth is based on knowledge and on the best estimation of uncertainties in risk assessment.

International cooperation is fundamental for identifying and advancing future nuclear safety research and applications and sharing investments. While CSNI constitutes a forum for international cooperation, the scope of activities is limited by resource availability, necessitating the exploration of new, more efficient collaborative models. Thus, to address future challenges and accelerate progress in accident analysis and management, innovative approaches to international collaboration is essential.

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This speech intends to cover four points: independence of regulatory authority, learning from accidents, regulatory challenges, and what we now expect from research.

Innovation requires strong and independent regulatory body. Regulation is often perceived as an obstacle to innovation. Many good operators and vendors may achieve the adequate level of safety even without any regulation, but the failure of a single poor competitor can drive all remaining technologies out of the market, taking an unreasonably long time and effort to recover from. There are numerous examples of conflicts of interest leading to poor decisions by organizations and their leaders. Maintaining the independence of regulatory authorities is essential for demonstrating and deploying advanced nuclear systems. Any erosion of regulatory independence puts the people and the environment at risk and significantly undermines public trust in nuclear technology.

Many people now seem to be trying to believe that severe accidents can be practically eliminated by design. However, new accident scenarios should be considered for new designs. There is still much to be learned from past accidents. The new design or feature or new practice shall also be adequately tested to the extent practicable before being brought into service, and shall be monitored in service to verify that the behavior of the plant is as expected.

There are numerous regulatory challenges, e.g., defining licensing basis events (LBEs). There are cases where the classification of states, such as normal operation, anticipated operational occurrences (AOOs), design basis events (DBEs), and design extension conditions (DECs), may need to be changed. There are also cases where the concepts of severe core damage or loss of containment function of specific barriers do not adequately describe the respective states.

For research, it is urgently needed to organize a framework to lead collaborative research projects with scaled experimental infrastructures to enhance development, validation and benchmarking of state-of-the-art codes, training and education.

This paper explores how artificial intelligence (AI)—particularly multimodal foundation models (FM), intelligent digital twins (IDT), and LLM-based multi-agent systems—is reshaping nuclear thermal-hydraulics and safety analysis (NTHSA). Historically grounded in physics-based modeling, structured validation, and expert-guided reasoning, NTHSA now faces growing demands for more adaptive, predictive, and transparent methodologies. Emerging AI technologies offer the potential to augment these foundations, enabling a shift from static safety analysis to a dynamic, epistemically intelligent safety paradigm.

The paper introduces foundation models—large-scale AI systems trained on diverse textual, numerical, and visual data—as tools that can reason, generalize, and automate complex tasks such as PIRT generation, closure model selection, and physics-code scripting. When embedded within intelligent digital twins, AI can enable real-time plant monitoring, anomaly diagnosis, and adaptive margin management, all grounded in both operational data and physics-based simulations. The integration of multi-agent architectures further allows the decomposition of safety analysis workflows into autonomous, collaborative AI roles—streamlining V&V, optimizing test matrices, and ensuring traceable, auditable recommendations.

This AI-enhanced framework not only accelerates traditional EMDAP loops but also opens the door to earning back conservatism through evidence-based learning. By dynamically reducing epistemic uncertainty over time, safety margins can be optimized while maintaining robust defense-in-depth. Case studies—such as the NAMAC framework and GPT-based discrepancy checkers—illustrate how AI can act as an assistant or advisor, improving explainability, trust, and operational awareness.

The paper also highlights critical challenges: limited nuclear data, explainability of black-box models, online V&V, cybersecurity, and human factors. It advocates for incremental adoption—starting with pilot deployments in non-critical systems, and expanding under transparent, auditable, and regulator-engaged oversight. Emphasizing ethics, human-AI collaboration, and sociotechnical integration, the paper charts a path toward AI as a trusted partner in nuclear safety.

Ultimately, AI is not portrayed as a silver bullet but as a transformative augmentation to the safety toolkit—empowering engineers and regulators to maintain high standards of performance and safety in an increasingly complex operational landscape.

The rapid advancement of artificial intelligence (AI) is reshaping global energy demand, notably increasing the need for stable, carbon-free power sources. As AI-driven services and data centers expand, major economies are revisiting nuclear power as a reliable energy solution. This paper analyzes recent discussions from NURETH-18 through 20, highlighting AI’s emerging role in nuclear thermal-hydraulics and system diagnostics. It further examines global nuclear expansion trends in response to projected electricity demand growth and decarbonization goals. The development of small modular reactors (SMRs), with enhanced safety and modular construction, is accelerating worldwide. Concurrently, regulatory frameworks are evolving to accommodate advanced reactor technologies, as exemplified by the U.S. 10 CFR Part 53 initiative. AI applications in nuclear operations, including anomaly detection, predictive maintenance, and documentation analysis, offer opportunities for efficiency and safety gains but raise new challenges in verification, regulation, and accountability. This paper addresses both technical and regulatory challenges for developers and regulators in adopting AI and deploying next-generation reactors. It concludes that while AI is driving power demand, it also holds potential to support nuclear innovation—provided appropriate safety, governance, and validation mechanisms are established.

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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 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.

This article presents key two-phase thermal-hydraulics research activities conducted at the POSEIDON and TOTEM platforms, located at CEA IRESNE in Cadarache. The facilities' capabilities support advancements in nuclear thermal-hydraulics research, with applications in reactor safety and innovation. Current studies include steam generator clogging in Pressurized Water Reactors (PWR) at the COLENTEC facility, aimed at enhancing maintenance and predicting clogging behavior. Passive safety systems development for Small Modular Reactors (SMR) is a significant focus, with tests at the EVEREST and EXOCET facilities evaluating natural circulation cooling efficiency. Research on compact steam generators at the MAGIC-3 and BICHE facilities targets performance improvements for space-efficient reactor designs. Additional studies on fuel cladding corrosion under PWR conditions at the CORAIL and CIRENE test loops contribute to more resilient cladding material development. Upcoming research will involve two-phase natural convection flows at the ANUBIS test rig to understand passive cooling in advanced reactors. The platform will also study subcooled boiling under PWR conditions at the DIOGEN facility to optimize heat transfer and safety margins. Lastly, in the continuity of studies for ASTRID demonstrator project, the PLATEAU and OLYMPE experimental loops could support R&D for Advanced Modular Reactors (AMR), including Sodium-cooled Fast neutron Reactors (SFR) and Molten Salt Reactors (MSR).

This study investigates the effect of condensation on the primary depressurization during a loss of coolant accident (LOCA) scenario in an integral small modular reactor (SMR). Two tests were conducted in an integral test facility, one with the passive residual heat removal (PRHR) system and break valve activated, and the other with only the break valve activated. Results show that condensation on the helical heat exchanger (HX) tubes has a significant impact on the primary system’s depressurization rate, which is found to be more important than the effect of the break itself. It is also observed that condensation water can compensate for the coolant loss caused by the break, leading to a slower decrease in the coolant level. The study highlights the importance of considering the effect of condensation in SMR LOCA accidents and suggests further research in this area.

Condensation is a key phenomenon for passive safety systems such as passive containment cooling system during an accident. Accordingly, numerical analysis tools are required to be sufficiently verified and validated for the development of a passive safety system. However, it is also challenging to predict condensation heat transfer precisely because various variables such as temperatures of wall and bulk fluids, geometric parameters, and non-condensable gas fraction affect the phenomena. In this study, we conducted condensation experiment and compared the numerical results from two one-dimensional system analysis codes with the experimental data. The condensation test facility was designed for simulating the condensation phenomenon in the small modular reactor passive safety system. Input models for the MARS-KS and SPACE codes were developed based on experimental facility geometric parameters and test conditions. Both MARS-KS and SPACE showed good agreements with experiment. However, SPACE code provides numerical option to choose model to calculate condensation. In other words, SPACE enables more detailed modeling than MARS-KS to choose an appropriate condensation model for the specific case. Accordingly, we investigated which model shows best agreement with the experiments and which model does not. These results suggest that selecting an appropriate condensation model according to the specific conditions of the condensation can enhance the accuracy of predictions.

Release of hydrogen in the containment of a nuclear power plant, during a postulated beyond design basic accident is a safety concern because explosive mixtures could form and damage components or even threaten containment integrity. The validation of computational tools against experimental data which a representative of postulated accident phenomena is an intermediate step aiming at enhancing the confidence in the code predictive capability. In this paper we present the experimental results of a series of experiments performed in the thermal-hydraulics PANDA facility investigating the erosion of a stratified containment atmosphere rich in helium (used to simulate hydrogen) by a vertical jet from a pipe, after interacting with a horizontal disk. For these experiments were used two PANDA interconnected vessels each of 4 m diameter and 8 m height (total volume 183.3 m³). The helium-rich layer was created in one vessel at the elevations 6 to 8 m. The jet was created by injecting steam from a vertical pipe with 20 cm exit diameter and 4 m elevation. The horizontal disk had a diameter of 20 cm, and it was installed at 5 m elevation. The experimental measurements include gas mixture temperature using thermocouples and concentration using mass spectrometer, and flow velocities using PIV. The tests with horizontal flow obstruction showed that decreasing the jet Reynolds number by a factor of two tripled the helium layer erosion time. On the other hand, changing the initial jet buoyancy does not have an appreciable effect on the overall helium layer erosion time.

The plate-type fuel assembly is widely utilized in nuclear research reactors and consists of several fuel plates and support plates. The fuel plate consists of fuel foil and metal cladding. The coolant is segmented into independent water gaps by the fuel plates and support plates. Due to the disturbances caused by the inlet structure of the plate-type fuel assembly, the flow velocity in each water gap is inconsistent. The significant differences in flow velocity between water gaps can lead to complex flow-induced vibrations in the fuel plates, potentially compromising structural stability. This study employs self-developed measurement technology to conduct detailed experimental research on the flow-induced vibration behavior of a simulated plate-type fuel assembly using strain gauges and eddy current sensors. The experimental results indicate significant differences in the deformation and vibration behaviors of the fuel plates along the axial direction. The deformation and vibration behaviors among the fuel plates also vary. The deformation at the inlet of the internal fuel plate is notably large. The deformation and amplitude at the entrance of the support plate are also notably large. However, the deformation at the outlet of the external fuel plate is larger. At low flow velocity, the amplitude in the middle axial region of the fuel plate is relatively large. At high flow velocity, the amplitude in the inlet region of the fuel plate is larger. The flow-induced vibrations at various positions of the plate-type fuel assembly do not exhibit a dominant frequency within the experimental flow velocity range.

The last Advanced Gas Reactor (AGR-5/6/7) experiment was conducted in the Advanced Test Reactor at Idaho National Laboratory from February 2018 to July 2020, accumulating 360.9 effective full power days. Since fuel temperatures could not be measured directly—because contact between a thermocouple and the fuel could lead to unwanted particle failures—the ABAQUS-based finite element heat transfer code was used to predict daily fuel temperatures over the entire irradiation period. Accurate determination of calculated temperature uncertainties is crucial in interpretation of fuel irradiation performance to ensure achievement of the AGR program objectives. Thermal model parameters with high sensitivity and/or large uncertainty were identified for quantification of the calculated temperature uncertainty. Propagation of model parameter uncertainty and sensitivity was then used to quantify the overall uncertainty of calculated temperatures. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish the sufficiency of the time-dependent first-order (linear) expansion terms in constructing the temperature response surface. Since heat produced in the fuel compacts is transferred through the gas gaps surrounding the compacts and graphite holder, uncertainty in the gap widths is a dominant factor in fuel temperature uncertainty. For all AGR-5/6/7 capsules, an error in capsule design allowed the graphite holders more lateral movement within the capsule shell than intended, resulting in a nonuniform gas gap around the capsule circumference that impacted fuel temperatures. This paper focuses on quantification of gap width uncertainties and the corresponding fuel temperature uncertainties during irradiation of the AGR-5/6/7 experiment.

In this study, a modified point reactor kinetics model is developed to account for the advection of delayed neutron precursors (DNPs) in a molten salt fast reactor (MSFR). Accurately capturing the behavior of delayed neutrons is crucial for MSFR transient analysis, as they have a significant impact on reactor control and overall stability. The point kinetics parameters, including prompt neutron generation time and the effective delayed neutron fraction, are calculated using an extended Monte Carlo N-Particle (MCNP) code. This version of the code is specifically modified to incorporate the effects of fuel circulation, which is a unique characteristic of molten salt reactors compared to traditional solid-fuel reactors. The modified model is implemented into the FLUENT using a user-defined function (UDF) to perform transient analyses for the unprotected loss of flow (ULOF) scenario. The reactor’s response to a sudden reduction in fuel flow is studied, focusing on how the core average temperature and reactor power evolve over time. The absence of recirculation zones in this transient scenario has significant effects on the inlet and outlet temperatures of the reactor, which are critical for evaluating the reactor's safety characteristics. The velocity and temperature fields within the reactor core during the ULOF event are analyzed in detail. The model is benchmarked against two independent models from the Politecnico di Milano and the Technical University of Delft, showing a good agreement with referenced results. This comparison validates the accuracy and reliability of the modified point reactor kinetics model for MSFR transient analysis.

In recent years, there has been renewed interest in Molten Salter Reactors (MSRs) for their potential advantages compared to reactors that rely on solid fuel. In response to such interest, many methods and codes have been developed to capture the unique features of MSRs. Among them, the System Analysis Module (SAM) is a modern system analysis tool that provides fast-running, modest-fidelity, whole-plant transient analyses capabilities, essential for fast-turnaround design scoping and engineering analyses of advanced reactor concepts. For liquid-fuel MSR, the complex physics and chemistry involved in MSR operation, such as reactor kinetics, fluid flow, heat transfer, and salt composition dynamics, pose significant challenges for system-level modeling. Specific modeling capabilities including are needed for system-level transient simulation. This paper presents recent advancements in SAM capability enhancements for system-level modeling of MSRs, focusing on improved simulation fidelity, computational efficiency, and multi-physics integration. Key enhancements include the development of species transport, Delayed Neutron Precursor (DNP) drift, modified Point Kinetics Equations (PKE), decay heat modeling, key fission product behavior, salt corrosion, and thermal-hydraulic coupling, as well as the code robustness and performance enhancements for MSR applications. The code enhancement allows for better predictive accuracy in safety analysis, transient behavior, and operational optimization, thus supporting the design and licensing of next-generation MSRs. Results from case studies are presented to demonstrate the benefits of these enhancements in accurately capturing key reactor transient behaviors.

Liquid-fueled molten salt reactors (MSRs) represent the only reactor design utilizing liquid nuclear fuel, wherein the molten fuel salt generates heat continuously during circulation, exhibiting unique fluid dynamics characterized by an embedded internal heat source. The presence of this internal heat source significantly influences wall heat transfer characteristics; however, experimental studies on molten salt heat transfer with internal heat sources remain scarce, leaving existing modified heat transfer models for fuel salts unvalidated by direct experimental evidence. This paper proposes an innovative experimental approach combining microwave heating and hot air heating to simultaneously simulate internal heat generation within molten salt and controlled wall heat flux. A rigorous calculation methodology for wall heat transfer coefficients under these coupled conditions is also established. The findings provide valuable insights and a methodological framework for experimental investigations of internal heat source-coupled heat transfer phenomena in liquid-fueled molten salt reactor systems.

In this study, the influence of internal heat sources on turbulent wall cooling heat transfer in a pipe was analyzed, targeting the heat exchangers in molten salt reactors (MSRs). The non-homogeneous problem arising from internal heat sources was solved using the superposition principle. Numerical calculations were performed from the entrance region to the fully developed region to account for the cumulative effects of internal heat generation. The local and mean Nusselt numbers (Nu) were calculated for a range of Reynolds numbers (Re) from 5 to 10⁶, Prandtl numbers (Pr) from 1 to 10, and internal heat source parameters (Ω) from 1 to 10³. The results indicate that the presence of internal heat sources under wall cooling conditions enhances the heat transfer rate. This enhancement becomes more pronounced with increasing Ω and decreasing Re and Pr. However, due to the thin viscous sublayer in turbulent flow, the maximum enhancement rate remains below 12%. Therefore, a region where an internal heat source produces a meaningful enhancement rate (≥ 5%) was identified. A correction factor was developed to account for the enhancement effect within this range. This study provides fundamental insights into the effects of internal heat sources and offers a quantitative basis for the design and performance evaluation of MSR heat exchangers.

This paper describes sensitivity analyses that were performed using the existing MSRE model for the STH code SPECTRA. The work described in this paper concentrated on the influence of uncertainties in salt properties on the heat transfer in MSRE during normal operation. The results of the current study lead to the following conclusions:

• Thermal conductivity has the highest (positive) effect on the heat transfer coefficient.
• Viscosity and specific heat have similar effect but with different sign; viscosity decreases the heat transfer coefficient, while specific heat increases it.
• The effect of density is smaller than the effect of thermal conductivity.
• Extended sensitivity analysis included the sensitivity indices from the Plischke method. It confirmed the strongest effect of thermal conductivity, followed by viscosity and specific heat (both having similar effect).
• Very similar results were obtained using both deterministic and random sampling methods.

This paper describes sensitivity analyses that were performed using the existing MSRE model for the STH code SPECTRA. The work described in this paper concentrated on the influence of uncertainties in the delayed neutron precursors (DNP) on the results of the MSRE low power transients: pump start-up and coastdown. The results of the current study lead to the following conclusions:

• There is a positive effect of the delayed neutron precursor yield fraction on the reactivity effect in both pump coastdown and pump startup transients.
• The effect is strong for the middle groups and weak for the long-living group and the short-living groups. The long-living group travels through the complete loop several times before it decays. The short-living groups decay mainly in the core, independently of the flow rate, and therefore these groups are basically not affecting the pump transients.
• Very similar results were obtained using both deterministic and random sampling methods.

In this study, a mass and energy (M/E) release analysis was performed on the postulated main steam line break (MSLB) accidents in the Advanced Power Reactor 1000 (APR1000). The M/E release rate was calculated using the SPACE-ME 1.0 code, developed by KEPCO Engineering & Construction Company, Inc. (KEPCO E&C), for various break areas ranging from an area fraction (AF) of 0.1 to 1.0, where AF 1.0 corresponds to the maximum double-ended guillotine break area. The initial core power was evaluated at 102%, 75%, 50%, 20%, and 0% of full power (%FP). To ensure conservative results, the break flow phase separation model and wall heat transfer multiplier were adopted. A simplified conservative model for the passive auxiliary feedwater system was used. The containment pressure and temperature responses were analyzed using the CAP 3.1 code with the calculated M/E release rates. A single failure of containment spray system was assumed. The highest containment peak pressure and temperature were found to be 0.7925 and 0.9531, respectively, which are normalized values with respect to the design values. The design margins of 20.75% for pressure and 4.69% for temperature during the most limiting MSLB accident indicate that the APR1000 containment can maintain its integrity well during the MSLB accidents. In conclusion, the new M/E release analysis methodology using SPACE-ME code is expected to be highly applicable to analyzing the postulated MSLB accidents in the APR1000.

This study presents the experimental results of large-scale containment thermal-hydraulics phenomena driven by the combined effects of steam injection and spray activation in a postulated Design Basis Accident (DBA) scenario, specifically a main steam line break. This experimental campaign, named P1A1_5 and P1A1_6, is part of the OECD/NEA PANDA project series. These experimental data could contribute to the assessment and validation of advanced computational tools for containment analysis.

The experiments were conducted in Vessel 1 of PANDA, a cylindrical confinement with 8 m in height and 4 m in diameter. A compartment representing the steam generator tower model was inside the Vessel 1. Initial conditions involved pressurizing with air and steam at 2.5 bar. There were defined with two different steam superheating (P1A1_5, P1A1_6), and the spray was activated using a single nozzle. The phases of the experiment were as follows: steam injection (phase 1), combined steam and spray injection (phase 2), and spray-only injection (phase 3 for P1A1_6).

Results showed that during containment depressurization, steam remains superheated above the spray nozzle. In contrast, below the spray nozzle, the fluid and saturation temperatures are approximately same value. Thus, the primary effects of steam injection and spray activation are the depressurization of Vessel 1 and the cooling of fluid and gas temperatures. Upon spray activation, the pressure and temperature gradients, especially below the spray nozzle, decrease more sharply over time compared to the phase 2. This is due to the enhanced momentum mixing and droplet behavior, particularly below the spray nozzle.

The NEA Committee on the Safety of Nuclear Installations (CSNI) has long supported international collaborations to enhance confidence in nuclear safety codes and experimental validation. One such initiative is the International Standard Problem (ISP), which began in the early 1970s and continues today.

A new ISP-52 was proposed upon the recommendations of the WGAMA/WGFS report: “Analyses of Design Extension Condition without Significant Fuel Degradation (DEC-A) for Operating Nuclear Power Plants” which highlighted the need for computer code validations for DEC-A conditions. For this purpose, the ETHARINUS project provided experimental data related to the PKL III J5.1 Run1 and Run2 tests. The latter addressed DEC-A scenario of Multiple Steam Generator Tube Rupture (MSGTR), which may occur following a severe earthquake, with limited safety system availability. In Run1 two double-ended guillotine breaks were considered in three out of four steam generators (SGs), while in Run2 the scenario was extended to all four SGs. Both test results were made available to the ISP-52 participants, but only Run 2 was selected for the “blind” and “open” analytical exercises.

This paper presents the main steps that have been followed to carry out the PKL III J5.1 Run1 and Run2 experiments and provide a description of the main events that took place during the course of the transient as well as the effectiveness of the operator actions (primary bleed, and manual activation of the ECCS) and the available ECCS to bring the system to safe shutdown conditions.

The TRACE code is used at Gösgen NPP to perform scoping simulations of plant behavior in response to extreme, very unlikely, external events. This allows plant personnel to investigate how the plant changes and refurbishments under study increase the safety margins in the event of Design Extension Condition (DEC), such as a Station Black-Out (SBO) with failure of the on-site emergency power supply. The safety of the unit is ensured thanks to the special emergency safety systems, bunkered and thus SBO-proven.

This study analyzes the benefits from the refurbished SEFW (Special Emergency Feedwater) tanks. In addition, this study investigates the automatic partial cooldown via Atmospheric Relief Valves (ARVs) (plant change not yet realized). Three simulations are presented in the paper. The full autarky time of 10 hours without operator actions is considered. In the first two cases the unit is kept at hot-shutdown conditions (no cooldown) and the secondary pressure is limited, respectively, by the cyclic opening of the Safety Relief Valves (SRVs) and the ARVs with automatic partial cooldown. The third case implements, based on the second case with ARV partial cooling, the manual cooldown procedure (slow gradient, 10 K/h) to reach cold shutdown.

The results of the simulations show that the automatic partial cooldown reduces the amount of primary coolant released into the containment by opening of the pressurizer safety valves. The increased safety margins of the plant in case of SBO are determined in 53 h (hot-shutdown state) and 42 h (cold-shutdown state) without external water injection.

During a Loss Of Coolant Accident (LOCA), in PWR’s, water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability. After the drainage of the Refueling Water Storage Tank (RWST), water is taken from sumps in the lower part of the reactor building. A filtering system is implemented to collect debris, such as fiberglass, paint and concrete particles, and to minimize the amount of debris entering in the core. IRSN has launched an experimental R&D project investigating the clogging of sump filters by integral tests performed in the VIKTORIA loop, which was equipped successively with two types of 2 m² filters used in 900MWe NPP’s. The debris carried to the filter generate at 80°C (with chemistry) a very quick increase of the pressure drop across the filter (≈ 1 to 7 kPa according to the debris source term) that could be due to rapid chemical effects further to fibers corrosion. The two types of filters (rectangular pockets or planar types) behave very differently with rather low head losses for the second type. The recent experiments performed with less amount of fibers (by replacement of part of fibrous materials by RMI metallic insulation) led to significantly reduce the head loss without any consequences on the downstream behavior (debris transferred to the core). The increase of the duration of the corrosion of zinc in acidic conditions (as a sensitivity study) lead to increase head losses by a factor 4 the which indicates the formation of chemical precipitates.

In the first decades of the construction of the nuclear reactors, the attention in the field of nuclear safety analyses was focused on major accidents starting from full power (with maximum initial energy in the system and decay heat). Later on, however, accidents occurring during shutdown were found to be as important as those occurring at full power. Abnormal operational events, postulated accidents and design extension conditions occurring during shutdown operational modes represent a significant contribution to the NPP risk due to the fact, that both preventive and mitigatory capabilities of the plant are partially or fully unavailable. Deactivation of safety features, equipment under maintenance or repair, reduced amount of coolant in some regimes, some instrumentation and measurements switched off or non-functionable; open primary circuit (loss of one barrier); and open containment (loss of another barrier) are the causes of the specific risk of accidents in the shutdown mode.

The core of the paper concentrates on the deterministic thermal-hydraulic (TH) safety analyses of the events starting from the shutdown operating modes of VVER-1000. Number of the performed analyses are long-term analyses specifying time windows for the operator (in situation with reduced availability of safety systems and their automatic actuation). Specification of VVER-1000 shutdown modes accompanied, availability of safety systems, methodology basis for the safety analyses, acceptance criteria, computer codes and their validation, list of scenarios analyzed for the VVER-1000, examples of analyses results, and incorporation of new analyses into Safety Analysis Report (SAR) will be described step by step.

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

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

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

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

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

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

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

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

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

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

SMRs and advanced reactors concepts can involve specific design characteristics requiring modelling capabilities that are beyond the validated boundaries of existing codes or include phenomena for which the existing experimental data is insufficient. The significant efforts and resources associated with performing validation or experimentation constitutes a challenge to a safe and secure timely deployment of SMRs. To overcome those challenges, the Internation Atomic Energy Agency (IAEA) set up a working group within the Nuclear Harmonization and Standardization Initiative (NHSI) to establish a Network for Experiments and Code Validation for Design and Safety Analysis Computer Codes for SMR and Advanced Reactor Designs (NEXSHARE). NEXSHARE is a technical forum of global cooperation and resource sharing for experiments and code validation between entities operating experimental facilities, design organizations of SMRs, Regulators’ Technical Support Organizations (TSOs) and other International Organizations. In particular, OECD Nuclear Energy Agency (NEA) and the Generation IV International Forum (GIF) are closely collaborating to this project. NEXSHARE was launched in 2024 at the IAEA Workshop on Experimental Testing and Validation for Design and Safety Analysis Computer Codes for SMRs. Feedback from the participants helped shape the next steps for the Network which include optimizing its functionalities, expanding its experimental facilities database, and conducting technology specific efforts on experiments and code validation. This paper provides an overview of NEXSHARE’s design and functionalities and also outlines the Network usages and benefits for the industry, supporting the IAEA’s initiatives to accelerate the development and deployment of safe and secure advanced reactors, including SMRs.

The McSAFER project was focused on experimental and numerical investigations for the safety evaluation of water-cooled SMRs such as NuScale, SMART, CAREM and F-SMR. The experimental program consisted in test series at three EU facilities e.g., MOTEL at LUT, HWAT at KTH and COSMOS-H at KIT. The experimental data was used for the validation of thermal hydraulic codes (CFD, subchannel and system codes). The experiments covered safety-relevant phenomena such as cross-flow in the core, the performance of the helical-coiled heat exchanger, forced and natural circulation and its transition, etc. The numerical part was devoted to the analysis of the core behavior under normal and accidental conditions (REA, Cold water injection) of four core designs (CAREM, NuScale, KSMR and F-SMR) using both industry-like and advanced transport Multiphysics computational routes. The behavior of a NuScale core loaded with ATF fuel under REA-conditions was investigated with three different high-fidelity coupling of neutronic, thermo-mechanics and thermal hydraulic codes and the obtained results were compared to the ones predicted for a core loaded with UO₂. Finally, selected transients (Steam Line Break for NuScale and SMART) were analyzed with three multiscale / multiphysics coupled codes including system TH, subchannel TH, CFD and 3D nodal diffusion codes. This paper will present and discuss the main outcomes of the core and plant analysis emphasizimg the capabilities and future improvements for a more realistic prediction of safety parameters of SMRs as well the potentials of the methods for the analysis of transients in nuclear power plants of Gen-2 and -3.

In the design and development of i-SMR, the passive emergency core cooling system (PECCS) is quite different from that of the existing reactors, and in particular, the depressurization valves and the recirculation valves may have completely different configurations and components from the existing ECCS valves.

The reason for such a complex configuration is that not only should the valves be able to be opened passively, but also actively opened by the actuation signals, and undesired opening should be prevented even with a single failure of the component.

Dynamic behavior of main valve consisting of spool discs, springs and orifices, block valve of specific shape, actuator trip valve and connecting pipes, etc., is critical at the validation of the design.

In this study, for this problem, the pressure and flow rate at each flowing part of the valve were calculated using the MARS-KS code, and the equations of motion of the spool disks were solved using the calculated flow data to determine the opening area of each valve, and the dynamic behavior was analyzed over time by feeding it back to the MARS-KS code calculation. The scheme was named as SEMICOM (Solution of Equation of Motion Implemented by Control-variables Of MARS code).

Using this method, it analyzes the dynamic response of a virtual PECCS valve and provides a requested performance data that can help determine various design parameters such as spring constant, disk-cylinder gap, and orifice size as well as dynamic stability determination.

The Horizon Euratom EASI-SMR project (Ensuring Assessment of Safety Innovations for light water SMR) aims to address critical R&D needs for the safety demonstration of Light Water (LW) Small Modular Reactor (SMR) technology, supporting its short-term deployment in Europe. Focusing on the European designs NUWARD and LDR-50, EASI-SMR targets innovations such as passive safety systems, boron-free cores, co-generation, additive manufacturing, and multi-unit operation. The project’s goal is to ensure that these reactors are designed, constructed, and licensed in accordance with European regulatory standards.

This paper discusses the core of the EASI-SMR project, which consists of three interconnected work packages: WP2 – Experimental Testing Program, WP3 – Code Validation and Scaling, and WP4 – Reliability of Passive Systems. WP2 establishes a new experimental program to investigate key physical phenomena in passive safety systems under both design basis and beyond design basis conditions, providing essential insights for LW-SMR safety demonstration. In WP3, the capability of European-developed codes to simulate DBA and BDBA scenarios is assessed, alongside the identification of best practices for passive system modeling and areas for code development. Finally, WP4 applies these validated codes to perform reliability assessments, focusing on risk analysis and licensing readiness for passive systems. This structured process, from experimentation in WP2 to code validation in WP3 and reliability assessment in WP4, creates a comprehensive and interconnected framework that addresses R&D needs, supporting the short-term deployment of LW-SMRs across Europe.

Most of the previously proposed data-driven fault diagnosis methods are Close Set Recognition (CSR) methods, which assumes that the training set and test set are drawn from the same fault label space. The resulting problem is that when facing an unknown type of fault that not included in the training set, CSR method will incorrectly classify it as one of the known fault types in the training set, bringing a huge negative impact on actual fault diagnosis tasks of nuclear power plants. The fault types of nuclear power plants cannot be exhaustive, and the fault types included in the training set are limited due to the difficulties in collecting and labelling data. Therefore, almost all nuclear power plant fault diagnosis tasks are essentially Open Set Recognition (OSR) tasks, which requires not only the correct classification of known fault types, but also the identification of unknown fault types. However, there are few related researches on OSR fault diagnosis in nuclear power plants.

To solve the above dilemma, a novel nuclear power plant OSR fault diagnosis framework based on CPL is proposed. Experimental data of 10 health states and 841 monitoring variables are generated by a detailed digital nuclear power plant model, which can truly reflect the high dimensionality and strong nonlinearity characteristics of nuclear power plant data. And 24 OSR tasks with different settings of known and unknown fault types are designed, on which the feasibility and effectiveness of the proposed framework are verified.

Heat pipe cooled reactors have garnered significant attention due to their simple design, scalability, and reliability, making them an ideal choice for nuclear power generation in space and deep-sea applications. The integration of a supercritical CO2 Brayton cycle system with heat pipes meets the demand for system miniaturization and high energy conversion efficiency in nuclear power systems. Although several conceptual designs have been proposed based on this idea, there is still a lack of research on the operational characteristics of these reactors, particularly concerning the impact of high-temperature heat pipe oscillations on system performance.

In this study, a coupled code combining a heat pipe-cooled reactor and a Brayton cycle system was developed to assess the transient effects of heat pipe temperature oscillations on system performance. The reactor code includes a neutron physics model, a heat pipe model, and a core heat transfer model, which were validated using reference data and experimental results. The supercritical carbon dioxide Brayton cycle system was modeled using a customized version of the Relap5 code, and the coupling between the two subsystems was successfully implemented.

Simulation results reveal that heat pipe temperature oscillations induce synchronous oscillations in the reactor and Brayton cycle system’s operational parameters, such as temperature and pressure. However, the operational state of the Brayton cycle system is less affected compared to that of the reactor system. This coupled code serves as an effective tool for the design and safety analysis of supercritical CO2 heat pipe cooled reactors.

This study, motivated by previous analyses from TEPSYS, investigates the impact of different cooling conditions on the distribution of non-condensable gases in the reactor building (R/B) of Fukushima Daiichi Unit 3 (1F3) during a severe accident. To understand this, experiments were conducted in the CIGMA facility, a large-scale test vessel that replicates the R/B structure. Steam and helium were continuously injected at the top of the vessel for 10,000 seconds to simulate steam and hydrogen leakage. CC-SJ-01, with a cooling temperature of 50°C, serves as the base case for comparison. In the present study, parametric investigations were performed under the same cooling conditions, focusing on the effects of increasing the partition aperture from one hole to nine holes (250 mm diameter each) and modifying the steam-to-helium mass ratio from 100:1 to 75:1. Results showed that the aperture change had little effect on helium distribution, with the highest concentration observed in the middle region, similar to CC-SJ-01. However, with the 75:1 steam-to-helium ratio, the highest helium concentration shifted to the upper region. The Shapiro ternary diagram revealed that a higher steam-to-helium ratio intersects the detonation limit in the middle region, while a lower ratio intersects it in the upper region. These findings are essential for understanding non-condensable gas behavior in severe accidents, aiding in the development of safety measures for nuclear reactor designs.

This paper presents a comparison between EDF MAAP 5.04 and EPRI MAAPv6.00 codes in simulating postulated Severe Accident (SA) scenarios in a generic integral PWR characterized by a submerged containment and about 60 MWe. This code comparison has been performed based on the Design 1 of the Horizon Euratom project SASPAM-SA (Safety Analysis of SMR with PAssive Mitigation strategies - SA).

The Modular Accident Analysis Program (MAAP) is a deterministic code developed by EPRI that can simulate the response of light water moderated nuclear power plants during accidental transients for Probabilistic Risk Analysis (PRA) applications. It can also simulate severe accident sequences, including actions taken as part of the Severe Accident Management Guidelines (SAMGs).

EPRI MAAP 5.04 does not natively enable to model iPWR transients: this code has been adapted by EDF (EDF MAAP 5.04) to make it compatible with the simulation of SAs transients for the SASPAM-SA Design 1. Conversely EPRI MAAPv6.00, the latest version of MAAP enables to natively model iPWR designs and the Design 1 of the SASPAM-SA project. EPRI MAAPv6.00 especially embeds new developments related to the In-Vessel Retention (IVR) evaluations, and support plate modeling that are similar to those implemented by EDF in the EDF MAAP5.04 version.

Comparisons performed between EDF MAAP5.04 and EPRI MAAP6 include accident progression from initial events to long-term in-vessel retention of the corium. Both Design Basis Accidents (DBAs) and SAs were considered. Particular attention was also paid to the solution adopted to reproduce the strong vessel-containment interaction typical of SMRs.

For nuclear disaster prevention, reviews of the Emergency Action Level during severe accidents are being examined based on lessons learned from accidents at Tokyo Electric Power Company Fukushima Daiichi Nuclear Power Station. In the examinations, it is important to consider various severe accident (SA) progressions, including very slow accident scenarios, and behaviors of fission product (FP) release. In this study, SA progression analyses of a typical Japanese 4-loop PWR were performed using the integrated severe accident analysis code MELCOR2.2, in the purpose of obtaining the knowledge utilized for the examinations of nuclear disaster prevention. In the analyses, the evaluation model of a typical Japanese 4-loop PWR was used, considering plant configurations, geometries and structural materials, countermeasure equipment and procedures against SA. SA progressions were compared among loss-of-coolant accidents with the different break sizes, such as guillotine-break, 6 inches-break and 2 inches-break of a hot-leg pipe. The results of the MELCOR 2.2 analyses showed that the speed of SA progression and the amount of FP released to the environment differed depending on the break size. It was also found that the FP releases increased in the late phase of SA progression, and their mechanism depended on the break size.

The Accident Source Term Evaluation Code (ASTEC) developed by IRSN has become a leading tool for the simulation of severe accidents in nuclear facilities. This mechanistic computer code models the entire accident sequence from initiating events to release of the source term outside the containment, including core degradation, containment behavior and fission product transport. Recent enhancements have significantly extended ASTEC's applications to Small Modular Reactors (SMRs) and their passive safety systems, Advanced Modular Reactors (AMRs), Accident Tolerant Fuels (ATFs), spent fuel pool accidents, potential incidents in fusion facilities such as ITER, and severe accident scenarios in fuel cycle facilities. ASTEC's capabilities are now extended to different reactor types, including Western PWRs, Russian VVERs, BWRs and CANDUs. It plays a crucial role in safety analyses, source term evaluations and the development of severe accident management procedures. The code is increasingly being adopted by research organizations, safety authorities and industrial companies for applications in existing and new reactor designs. ASTEC supports probabilistic safety assessments, emergency preparedness and interpretation of experimental programs. The flexibility of the software has facilitated its integration into the new European project ASSAS, which focuses on the use of artificial intelligence for severe accident simulation. This paper provides an overview of the new applications of ASTEC in nuclear reactor simulation and related R&D activities. It highlights the importance of the code in improving nuclear safety assessments and its integration into international projects on advanced nuclear technologies, including European initiatives focused on SMRs and passive safety systems.

Steam explosions represent a critical challenge in the management of severe nuclear reactor accidents, particularly in Nordic Boiling Water Reactors (BWRs), where unique operational and environmental conditions affect accident progression. This study focuses on analyzing steam explosions during severe accidents in Nordic BWRs using a coupled MELCOR-TEXAS computational framework. MELCOR, a widely used code for severe accident analysis, provides detailed thermal-hydraulic and fission product behavior modeling, while TEXAS specializes in simulating fuel-coolant interactions and steam explosion dynamics. By coupling these codes, we achieve a comprehensive simulation environment to evaluate steam explosion scenarios with a higher level of accuracy.

The research investigates core melt progression, melt relocation to the lower plenum, and the conditions leading to steam explosions upon interaction with coolant water. Key parameters assessed include melt jet breakup, premixing dynamics, vapor film stability, and pressure wave generation. The study emphasizes factors influenced by Nordic BWR design features, such as high-density containment structures and emergency core cooling systems. The coupled MELCOR-TEXAS model enables a detailed examination of pressure loads on containment structures, critical for understanding potential damage thresholds.

Results from these simulations enhance our understanding of steam explosion risks specific to Nordic BWRs and support improvements in accident management and containment design. Findings from this work aim to inform safety guidelines and regulatory standards, contributing to robust safety measures in the context of Nordic nuclear facilities and advancing preparedness for severe accident scenarios.

An essential safety measure of advanced water-cooled nuclear power plant designs is the use of passive safety systems (on safety level 3) for the control of design basis accidents (e.g. cooling condensers) or of special devices on safety level 4 for the prevention and mitigation of severe accidents (e.g. ex-vessel core catcher concepts). GRS is developing the code package AC² for simulation of safety relevant phenomena and processes from the initiating event up to the release of fission products to the environment. AC² consists of ATHLET for the simulation of the reactor cooling system, CD for the core degradation phenomena and fission product behaviour in the coolant circuit and COCOSYS for the simulation of all phenomena describing the containment thermal-hydraulic state and potential fission product release to the environment in case of severe accidents.

A key consideration in developing integral simulation codes such as AC² is determining the appropriate level of detail needed to accurately represent all essential processes while ensuring the calculation time remains manageable. This aspect is addressed in this paper with a focus on selected passive systems/devices that found application in Generation III+ reactor concepts, such as the Russian type WWER-1200 light water reactor: Passive heat removal systems and ex-vessel core catcher devices. AC² provides new modelling features for their simulation based on an improved coupling between ATHLET/CD and COCOSYS. The basics of the new COCOSYS modelling concepts for passive systems are described and the paper shows their combined performance in a single calculation for a WWER-1200.

This paper presents a summary of the results from phase 1 of the Critical Heat Flux (CHF) benchmark, organized by the OECD/NEA Task Force on Artificial Intelligence (AI) and Machine Learning (ML) for Scientific Computing in Nuclear Engineering. As the first in a series of AI/ML-focused initiatives led by the Task Force, this benchmark received 48 contributions from 31 institutions across 14 countries. Phase 1 focused on CHF regression in uniformly heated vertical tubes, utilizing a large public database of 24,579 CHF data points and a separate blind database containing 560 CHF points. Most submitted models employed either neural network architectures or gradient-boosting decision tree methods. Independent evaluations of the model predictions were conducted using standard statistical metrics applied to both databases. Model overfitting, generalization, and extrapolation performance were assessed using predictions for the blind database and various slice datasets. The results indicate that most ML models significantly outperform the 2006 Groeneveld CHF lookup table by at least a factor of 2, in part due to the explicit consideration of the heated length effects. For the considered databases, tree-based methods demonstrated superior performance, including in extrapolation scenarios to large tube diameters and high mass fluxes. While model accuracy generally improved with increasing model size (in terms of number of trainable parameters), a few promising models achieved high accuracy while maintaining a reasonable size.

Critical heat flux prediction models in water-cooled systems often rely on local equilibrium quality to avoid tracking upstream history effects. Historical critical heat flux experiments with uniformly heated vertical tubes suggest that, in sufficiently long tubes, outlet quality is an adequate parameter to substitute for the combined dependence on heated length and inlet temperature.

In this study, machine learning techniques were applied to the critical heat flux database of the United States Nuclear Regulatory Commission---the foundation for the 2006 critical heat flux look-up table of Groeneveld---to assess input parameter importance and redundancy using game-theoretic Shapley values. Supported by this analysis, three machine learning models were trained for critical heat flux prediction. Each model employed the same machine learning technique and four common input parameters: pressure, mass flow, tube diameter, and heated length. The fifth parameter varied between outlet quality, inlet temperature, and inlet subcooling.

The results confirmed previous findings that replacing parameters defining outlet conditions with those describing inlet conditions improves the statistical performance of critical heat flux prediction models. Additionally, using subcooling instead of temperature enhances predictive accuracy, particularly in cases where phase change is already occurring at the inlet.

Critical Heat Flux (CHF) plays a pivotal role in ensuring reliability and safety within boiling two-phase flow systems. Despite the development of numerous CHF prediction tools using conventional empirical correlations, machine learning, and deep learning methods, the complex mechanisms underlying CHF continue to challenge the development of a unified, accurate, and robust prediction model. The complexity is further exacerbated by varying experimental dataset developed over the decades of CHF research. In response to these challenges, the present study leverages state-of-the-art AI method, including ANN, CNN, Transformer model, and transfer learning techniques. The proposed AI-based CHF prediction model, particularly the Transformer model employing self-attention mechanisms, dynamically assigns importance to different parts of the input data. The approach significantly improves the model's capability for CHF prediction. The results indicate that the predictive performance of the Transformer-based AI model exceeds that of the Look-Up Table (LUT) method and a benchmark model from the OECD-NEA based on the database encompasses 24,579 CHF data point conducted in vertical, uniformly heated, water-cooled tubes from 59 distinct sources over the past 60 years. The five-input AI model achieved the best predictive performance: Mean P/M of 1.008, Std. P/M of 0.122, RMSPE of 12.3%, MAPE of 7.22%, NRMSE of 9.91%, and Q2 of 1.26%. Moreover, the AI-based CHF prediction model's prediction behaviors are examined and compared with the LUT method. This comparison confirms the model's resistance to overfitting. Finally, by utilizing transfer learning, the model's ability to predict CHF in tubes is extended to annulus and plate geometries.

The prediction of the Critical Heat Flux in a light-water nuclear reactor core is crucial for design, operation and safety. One of the most successful method to predict the CHF is to use the Groeneveld Look-up Table. It consists of the interpolation of more than 25000 CHF experimental data in tube geometry, recently collected in a dedicated NRC database. However, its accuracy is not completely satisfactory. To better understand and predict the CHF, the OECD/NEA Expert Group on Reactor Systems Multi-Physics (EGMUP) mandated a new task force for the development of an artificial intelligence strategy for CHF regression. In this context, the present article describes a series of machine-learning regressions applied to the NRC database to predict CHF. Trying with an increasing-complexity approach, the technics of SVR, Multilayer Perceptron, Physics-Informed Neural Network, Residual Network are applied. Almost each ML-technic gives better results than the Look-Up Table. The most performing one is the Residual Network 30x64 with a RMSPE of 11%. A sensitivity to feature selection, such as mass flow rate, pressure, diameter, length of the tube, inlet temperature, and enthalpy is performed. The length of the tube is decisive to have a good accuracy even if its physical role in the prediction is debatable.

Critical heat flux (CHF) is one of the most important parameters in the design and safety analysis of water-cooled reactors. In the past, extensive experimental studies were carried out by various researchers to understand the physical processes and to provide experimental data bases for the development of prediction models. Due to some practical reasons, such as privacy issue, experimental data obtained by one researcher cannot be made available for other researchers. This led to hundreds of prediction models with narrow valid parameter ranges.

Recently, machine learning (ML) method has attracted more and more interests in the CHF prediction. The necessary condition for a successful ML-model is a large data base for training and testing. The present study proposes a new method, called data-informed continuous machine learning (DI-CML), with the key feature to generate an artificial data base, which is almost similar to the data base of the previous researchers without knowing any original experimental data points.

This paper describes briefly the idea of the DI-CML approach, which is applied to the CHF prediction with the large CHF data base provided by the OECD-NEA benchmark working group. The results achieved so far confirm the feasibility of the DI-CML approach. At the same time, challenging issues and open tasks for the future research works are pointed out, to further develop and to improve the DI-CML approach.

New small modular reactor technologies are being developed, having in common innovative compact design and a reliance on passive safety systems for enhanced safety. In the framework of the H2020 European project ELSMOR, a generic pressurised water compact design was defined for safety study purposes. A model for this study design has been built with the ASTEC V3.1.2 code and includes a passive heat removal system loop (PHRS) that should ensure the extraction of the reactor residual heat during an accident. The plant response during a station black-out scenario has been investigated when the safety system is active. While the system is able to extract the heat and keep the core covered, pressure in the loop is shown to be highly dependent on the modelling of the upper plenum area. During the PHRS operation, flow instabilities could be observed in the reactor primary loop. The mechanisms leading to the triggering and stopping of these oscillations in the calculated flow are analysed. The sensitivity of the global plant behaviour to different geometrical parameters of the safety system such as pipes diameter is also studied. As long as the PHRS can extract the residual heat, a very similar plant response is observed. The last investigated parameter is the passive loop filling ratio. This parameter is shown to have only a small impact on the heat removal capacity of the loop but can influence oscillatory flow development in the PHRS secondary loop.

Passive systems are being considered for advanced reactor designs owing to their enhanced reliability against an extended loss of onsite power. Particularly, the Safety Condenser (SACO) stands out because of its capacity of passively removing core decay heat through the steam generators by condensing steam inside a heat exchanger immersed in an external water pool. This work, embedded within the PASTELS European Project, presents experimental and numerical results on SACO performance at the PKL facility.

The data obtained in this study concerns a vertical straight-tube SACO immersed in a water pool. The SACO is connected to the secondary side of the PKL facility through charge and return lines. The P1.2 test consist of several secondary-side depressurization sequences aiming at studying the dependency of SACO power removal on secondary side pressure and nitrogen initial mass, pool water temperature and straight-tube liquid level. These results are then compared to the predictions of the CATHARE-3, ATHLET, TRACE and RELAP-5 system thermalhydraulic codes.

Experimental results on P1.2 experiments show the possibility to limit the SACO power removal by controlling the fill level in the SACO straight tubes, which is important for accident management purposes. Concerning simulation results, system thermalhydraulic codes generally overestimate depressurisation rates. In some cases, deviations can be ascribed to an inadequate modelling of a specific component (e.g. auxiliary heater, nitrogen injection), while in other cases they are related to the modelling of the pool and the difficulty of capturing the redistribution of nitrogen within the straight tubes along the transient.

Passive safety systems are increasingly utilized in prospective nuclear power plant designs. The low magnitude of the forces involved in such systems, combined with the uncertainty inherent in the factors affecting them, poses a problem in assessing their reliability compared to active counterparts.

The purpose of this paper is to investigate and apply a state-of-the-art technique in passive reliability assessment, known as the Reliability Methods of Passive Systems (RMPS) methodology, to the isolation condenser system (ICS) of the prospective BWRX-300 small modular reactor (SMR) design. The ICS is a safety system driven by natural circulation that provides emergency core cooling and pressure control for the BWRX-300. Using RMPS to analyze the effect of uncertainties in (a) the thermal characteristics of the fuel and (b) two-phase constitutive correlation factors on ICS operation, the reliability of natural circulation was quantified with a confidence of 95%, yielding an immeasurably small failure probability. Considering residual uncertainty, an engineering judgment assigned a failure probability of 1.00E-07. This finding was integrated into a fault tree analysis of the ICS using failure mode and effect analysis (FMEA) of system components, including insufficient natural circulation as a failure mode. Analysis of sequences leading to failure resulted in system unavailability being determined as 1.62E-07 for the case of all three loops initially available and 2.91E-05 for the case when only two loops are initially available. Sensitivity analysis of the natural circulation failure probability with respect to ICS system unavailability was also performed to investigate the robustness of the design.

A numerical simulation study of a Two-Phase Loop Thermosiphon (TPLT) in a novel Small Modular Integrated Heating Reactor (SMIHR) is conducted. First, a comparative analysis of TPLT performance with different design parameters is performed to determine the baseline parameters. Then, the phase change, natural circulation, and heat transfer characteristics of the TPLT under various operating conditions are investigated. The results revealed a complex relationship between these parameters and the performance of the TPLT. These insights provide valuable guidance for the design and optimization of TPLTs.

Computational Fluid Dynamics (CFD) is widely used for thermohydraulic problems, with the RANS (Reynolds-Averaged Navier-Stokes) approach being popular due to its fairly good quality/cost compromise. However, unsteady complex phenomena such as fluid structure interactions (FSI) or thermal fatigue cannot be predicted with a RANS or even (U)RANS (U: Unsteady) simulation. With the growth of computing resources, Large Eddy Simulation (LES) is increasingly used to model this kind of phenomena, even for high Reynolds number flows that require modeling at the walls (WM-LES: Wall-Modeled LES).

The present communication exhibits an update on validation test-cases computed using the EDF’s open-source code_saturne V8.0 software. These cases include fully turbulent pipe flow, an impinging jet on a heated plate, a wall mounted cube, a backward facing step and a 90 degrees bend pipe. They may be encountered in several location in the primary circuit and the validation on this non-exhaustive list of test-cases is crucial for every approach such as WM-LES or zonal and non-zonal hybrid RANS/LES before applying it on industrial applications. The simulations use the LES approach with standard numerical options, and the standard Smagorinsky sub-grid scale model. Comparisons are made on quantities such as the velocity field, the Reynolds stress tensor and the Nusselt number. WM-LES results show fairly good agreement with experimental and DNS data, particularly for the mean velocity field. The turbulence might be well predicted but remains a challenging issue when the wall is modeled with a standard wall function.

In the context of nuclear engineering, Flow-Induced Vibration (FIV) in steam generators can lead to mechanical damage, responsible for safety issues and significant maintenance cost in Nuclear Power Plants (NPPs). Before going towards FIV simulations with moving tubes, validating tube bundle simulations with fixed tubes is needed.

The present work is performed in the framework of the GO-VIKING Euratom European project (Gathering expertise On Vibration ImpaKt In Nuclear power Generation, https://go-viking.eu/). The open source CFD solver code_saturne (www.code-saturne.org) developed by EDF is utilized using massive parallel computing.

Wall-Resolved LES (WR-LES) is first revisited and validated for the flow around a cylinder at an incident Reynolds number of 3900 with 0% inlet turbulence and a periodic boundary condition in the spanwise direction. Several refinements and computational domains are used and conclusions are drawn to correctly predict drag and lift coefficient for this flow with the actual discretization scheme.

After simulating the flow around a single cylinder with a non-zero incident turbulence, the flow through a 3.5 x 5 tube bundle configuration with a pitch-to-diameter ration equal to 1.44 is studied. AMOVI CEA experimental data are used for comparisons. No periodic boundary conditions are employed in the span-wise direction, the full experiment test-section being represented with a no-slip boundary condition for all the walls. A particular attention is given to comparisons of the drag and lift spectra in time between the experiment and the CFD results.

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.

A coupled neutronics and thermal-hydraulics simulation code is developed at JAEA. In the coupling simulation code, the 3-dimensional two-fluid ACE-3D code, which is the in-house code of JAEA, will be adopted to simulate thermal-hydraulics behavior inside nuclear reactor fuel assemblies. The ACE-3D code calculates void fraction distributions under operational conditions for use in neutron transport simulations. This research aims to validate ACE-3D using data from a two-phase flow experiment. For this purpose, a two-phase flow experiment was conducted in a 4×4 unheated fuel assembly. In the experiment, the time-averaged void fraction distribution was measured using a wire mesh sensor system under high temperatures (373 K-500 K) and high-pressure conditions of up to 2.6 MPa. The experimental results were analyzed, and the data were visualized to understand better the behavior and characteristics of the two-phase flow in the fuel assembly. A two-phase flow data set is being developed, covering a wide range of experimental conditions, including higher-pressure regions, which can be used for validating thermal-hydraulic codes. Finally, the ACE-3D thermal hydraulics code was applied to the two-phase flow experiment. The calculation results were then compared to the experimental ones, and the issues were identified for improving ACE-3D in future simulations.

CFD is considered to be a valuable tool for assessing and improving the performance and safety of nuclear reactors. Verifying or creating CFD models to predict reactor fluid dynamics is crucial for Gen-IV reactors, which are cooled by liquid metals. The thermal boundary layer in liquid metals is of a greater thickness than that of the momentum layer, leading common turbulence models to incorrect predictions: this highlights not only the necessity for the development of new models but also the creation of databases to validate them. Two main methods may be used to collect the data: perform experimental tests or conduct high-fidelity simulations. This study compares these two methods by assesing the results of a benchmark exercise proposed by the EGTHM. The reference system for the thermo-hydraulic exercise is the Advanced LFR European Demonstrator. Particle Image Velocimetry was employed to obtain high-resolution data on the flow around a triangular lattice.* Computational high-fidelity data are obtained via DNS using an original discretisation technique on a periodic domain of four subchannels. The numerical study also considers heat transfer, by setting a Prandtl number Pr=0.031 representative of LBE. Statistics of velocity, thermal fields and main turbulent flow features are presented and compared with experimental data. This approach allows not only the comparison of experimental and numerical results, but also the integration of one with the other where there are deficiencies, with the aim of providing the optimal dataset for the development of future turbulence and heat transfer models.

*Menezes et al., DOI: https://doi.org/10.1063/5.0154590

Spacer grids in PWR fuel assemblies are equipped with mixing vanes for inducing lateral coolant flow and improving thermal performance. Therefore, understanding how shape of mixing vanes has impacts on flow is important in spacer grid design . This paper presents the results of the PIV (Particle Image Velocimetry) experiment, which was performed on 5x5 rod bundles with spacer grids of two different designs. From the results, the shape effects of mixing vanes were evaluated. Additionally, the experimental results were compared to the results of CFD (Computational Fluid Dynamics) and the applicability of CFD in spacer grid design was investigated.

This study presents a novel application of Lagrangian Particle Tracking (LPT) combined with plenoptic imaging to perform three-dimensional flow measurements within a 6x6 nuclear fuel rod bundle. The experiments were conducted in the Shaking Bundle Facility (SBF), a full-scale experimental model of a nuclear fuel assembly featuring 36 acrylic rods arranged in a square lattice. The assembly replicates the rod-to-rod pitch and mechanical properties of prototypical fuel bundles, with spacer grids providing structural support and simulating realistic flow conditions. Para-cymene, a working fluid with the same refractive index as the surrogate rods, was used to achieve optical transparency and accurate flow measurements.
The experimental setup enables detailed characterization of mean flow fields, boundary layer profiles, and turbulence statistics. Flow data were captured using a plenoptic 3.0 imaging system capable of volumetric measurements in a 14x12x12mm domain, with particles tracked at a frequency of 2 kHz. Advanced Radial Basis Function (RBF) interpolation was applied to reconstruct smooth 3D velocity fields while enforcing incompressibility and boundary conditions.
Results are systematically compared to Direct Numerical Simulations (DNS), Large Eddy Simulations (LES), and prior experimental data, providing a robust benchmark for validating thermal-hydraulic models. This work offers a high-fidelity experimental dataset for improving the predictive accuracy of computational models used in nuclear reactor safety analysis and for understanding flow behaviors under reactor-like conditions.

The depressurization process in light water reactors is an important factor for nuclear safety, and there is a need to develop an analysis code for this transient phenomenon and its validation process. Corroborated experimental data are crucial for evaluating the thermal characteristics of transient boiling and its associated uncertainties. In particular, the spatiotemporal distribution of void fraction during the depressurization process remains undetermined. This study conducted a transient flow boiling experiment during a depressurization process with our test facility for 3D thermal hydraulics in light water reactors (SIRIUS-3D). The test section was a 5×5 rod bundle partially simulating the fuel assembly of an actual reactor. Five units of the subchannel void sensor, capable of measuring the local void fraction between electrodes at a high sampling rate, were installed along the axial direction in the test section’s heated region. We evaluated the multi-dimensional void behavior in a 5×5 rod bundle with a linear depressurization rate ranging from 0.5 to 2.0 MPa/s and an initial system pressure of 7.2 MPa. The rod-surface heat flux and inlet mass flux were set to 70 kW/m² and 750 kg/m²/s, respectively, for all cases. The development of the boiling flow during the depressurization process was summarized with the depressurization rate as a parameter. The void fraction growth rate and time-averaged void fraction were quantified. The spatial void fraction distribution was organized and discussed based on the average values obtained by dividing the regions according to the distance from the center of the bundle cross-section.

The United States has the largest operating fleet of nuclear reactors in the world. Operating cost reduction and power uprates are the two major topics that can further bolster the economic and technical sustainability of LWRs. Over the past decade, a large amount of two-phase flow and heat transfer data, especially for the post-CHF regime, has been collected through the NRC/PSU RBHT reflood test facility. The present work summarizes the on-going efforts in large-scale RBHT data compilation and the assessment of heat transfer modeling accuracy for the dispersed flow film boiling and the transition boiling regime under elevated pressure conditions (up to 60 psi). Various existing post-CHF heat transfer models were compared with the rigorously compiled large-scale experimental data sets, and their performances were evaluated. It was found that existing models carry with them large prediction uncertainties, which further leads to an overly conservative safety limit for existing LWRs. Therefore, it is recommended to develop new post-CHF heat transfer models with significantly reduced thermal hydraulic uncertainties that originate from advanced two-phase flow diagnostic and processing data techniques as well as more in-depth understanding of the underlying physics involved. The large-scale dataset is also found to be extremely useful for developing physic-integrated data-driven models that provide superior prediction accuracy with physically realistic projections.

In validation exercises, numerical simulations are compared directly with experimental data. If the agreement is sufficiently good, then the model is considered validated (with the reported accuracy) and it can be used with a high level of confidence for the investigation of closely related scenarios that have not been or cannot be studied experimentally. A key element in this comparison is the justification of the modeling assumptions and simplifications, demonstrating that the dominant physical phenomena are well represented, and thus simulations of similar scenarios are reliable.

Rod bundles with wire spacers are used as fuel assemblies in many liquid-metal cooled fast reactor designs. The thermal-hydraulic scenario is complex due to the intricate geometry and the low Prandtl number of the coolant. While several experimental campaigns are reported in the open literature, some important aspects must be taken into account for comparing numerical and experimental results. This work discusses the impact of geometry simplification (e.g. wire shape) and uncertainty in the location of thermo-couples, as well the influence of material properties and boundary conditions. Moreover, relative errors must be defined with respect to the correct reference value in order for them to strongly support conclusions regarding the accuracy.

Two reference cases are selected for detailed analysis. They cover the study of the velocity profile in an isothermal case, and the temperature profile in a heated case with local blockages.

2.14.0.0

Liquid Metal Fast Reactor (LMFR) rod bundles can be designed with grid spacers or wire wraps. In both types of designs, pins may deform due to tension of the pre-stressed wires, contact pressure between clad and adjacent rods and/or wires, thermal and irradiation clad creep, irradiation-caused swelling and fuel burnup. In order to analyze the effect of such deformations on the peak temperature and temperature distribution, and in order to validate a simulation methodology, a series of experiments in water and liquid metal for grid-spaced and wire-wrapped liquid metal cooled fast reactor rod bundles is being conducted in Europe. These experiments are supported by numerical analyses which will be described.

Experiments in water bundles aim at validating the simulated flow field around a bended pin, while at the same time giving a first impression on the validation of the temperature field. Similar experiments in liquid metal rod bundles aim to validate the temperature field in the simulations.

A description of the water and liquid metal experiments will be provided and subsequently the numerical support to these experiments will be discussed completed by a future outlook.

Lead-bismuth Eutectic(LBE) is a excellent coolant for the Small-Modular Reactor(SMR). And there is a Plate-type fuel assembly is considered in our research, making full use of its advantages of tight structure and high heat transfer efficiency will provide a wide prospective for the development of SMR. The comparable research about the convective heat transfer characteristics of LBE in horizontal and vertical rectangular channels is conducted in this paper. There are 9 subchannels in a assembly and the aspect ratio of each subchannel is about 21.4. The flow rate distribution characteristics and the convective heat transfers characteristics are compared and analyzed. Within the range of 370000 ≤ Re ≤650000 and 300℃ ≤ T ≤450℃, the convective heat transfer mechanism is researched. The critical working conditions for this structure are proposed for the natural convection, mixed convection and forced convection using the Large Eddy Simulation(LES) model, and the influence of buoyancy on the turbulent heat transfer process is also analyzed.

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics are analyzed with the sub-channel analysis method. The effects of different blockage areas and axial positions are considered. The results indicate that when a flow blockage accident occurs, larger blockage areas and blockage positions closer to the axial center result in more severe accident consequences. However, for the blockage scenarios studied in this study, all peak temperatures remain below the material limit temperatures. This work could provide a reference for the future design and development of the LFR.

Liquid metal cooled fast reactor (LMFR) designs typically utilize helically wire-wrapped pin bundles. The high thermal conductivity of the liquid metal, combined with the thermal resistance of the wire-wrapped contact points, leads to localized hot spots under high heat flux conditions characteristic of the LMFR. A separate effects test was conducted under stagnant fluid conditions to measure the local temperature peaks at the wire contact point using an infrared thermography. Galinstan was selected as a primary test fluid to simulate the thermal hydraulics characteristics of the low Prandtl number fluids in LMFRs and the analysis of local hot spots was performed with the various wire configurations and fluids. In addition to the experimental investigations, the study of hot spots was further expanded to the forced convection conditions through computational fluid dynamics (CFD) simulation, allowing for a more comprehensive analysis of the effect of the influencing parameters. Hot spots were observed when the Biot number exceeds unity, exhibiting a direct proportional relationship with the local heat flux. Based on both experimental data and CFD-generated data, an empirical correlation was proposed to predict the severity of the hot spots, which can be applied across different material and flow conditions. This study provides valuable engineering insights and recommendations for wire design strategies aimed at mitigating the undesired hot spots by reducing the contact area and enhancing the heat transfer coefficient.

Lead-bismuth reactors, with their advantages of high power density, strong inherent safety, and good maneuverability, have received widespread attention. Based on the subchannel model, a thermal-hydraulic model for a small lead-bismuth reactor assembly was established. Using a statistical analysis framework, Latin Hypercube Sampling was adopted as the sampling method, while the Wilks method and Spearman method were used for tolerance interval estimation and sensitivity analysis, respectively, to develop an uncertainty analysis program for lead-bismuth subchannels. Referring to pressurized water reactors, the selected input parameters include coolant inlet flow rate, inlet temperature, outlet pressure, power, fuel thermal conductivity, and mixing coefficient. The chosen output parameters are the average coolant temperature, the coolant temperatures in different types of channels, and the surface temperatures of different types of fuel rods. A total of 5000 samples were drawn for uncertainty analysis. The results indicate that the tolerance interval range for the average coolant temperature is smaller than that for the subchannels. The mixing coefficient has a lesser impact on the former but a greater influence on the coolant temperatures at different positions and the fuel rod temperatures.

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

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

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

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

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

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

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

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

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

CINEMA (Code for INtegrated severe accident Evaluation and Management), a severe accident analysis code developed in South Korea, consists of several modules enabling independent analysis of complex severe accident phenomena. Previous validations studies, including simulations of the Three Mile Island (TMI) accident and a comparison with the MAAP, demonstrated that CINEMA simulated both thermal hydraulic behavior and accident progression well. Further comparative analyses with other well-validated codes that cover both accident sequences and fission product (FP) behavior can enhance understanding of CINEMA’s simulation characteristics. This study compares the results of MELCOR and CINEMA’s analyses of accident progression and FP behavior during SBLOCA (Small Break Loss Of Coolant Accident) for OPR1000. Both codes showed broadly consistent sequences of major events, and CINEMA’s detailed NSSS nodalization showed specific fluid flows. Regarding FP transport, both codes predicted the inert gas Xe mostly remained suspended in the containment building. For Cs and CsI, both codes assessed that these species did not exist in an airborne with in the containment building. However, MELCOR predicted that the release fraction to the containment building could vary with break size. In CINEMA, Cs and CsI were deposited more within the reactor coolant system than in the containment building. The FP behavior was influenced by the flow direction in the RCS and modeling such as chemisorption, pool deposition, and re-evaporation.

This paper presents an experimental study on the pressure drop characteristics in fixed beds packed with sand particles, with the goal of improving the accuracy of pressure drop predictions. Single- and two-phase flow tests were conducted using a custom-designed, adiabatic test facility specifically built to investigate the frictional behavior of flow through porous media. The facility allows for precise control and measurement of flow conditions, providing robust data for analysis. Using the effective diameter derived from single-phase flow tests in sand-packed beds, two-phase flow experiments were performed, and various prediction models were validated by comparing the measured pressure drop data against calculations from different analytical models. The results demonstrate that for two-phase flow in beds packed with smaller sand particles, the measured pressure drops increase steadily with fluid flow rate. In contrast, for beds with larger, coarser sand particles, the pressure drops exhibit an initial decrease followed by an increase as flow rate rises a down-up tendency. Notably, only models that account for interfacial drag effects successfully predicted this behavior. However, despite this, the prediction models showed significant deviations from the experimentally observed data, highlighting the complexity of accurately modeling two-phase flow in porous media. These findings suggest the need for further refinement of predictive models to better capture the intricate behavior of two-phase flow in such systems.

Various types of non-light water reactors are currently engaged in the U.S. licensing process. Because of inherent differences compared with well-established large light water reactors, appropriate assessment tools are needed. Specifically, source term analysis, which determines environmental dose impacts from potential accident scenarios, is a crucial part of design and licensing. The U.S. Nuclear Regulatory Commission has emphasized the importance of mechanistic source term analysis for advanced reactor deployments. To align with these needs, Argonne National Laboratory has developed the Simplified Radionuclide Transport (SRT) source term analysis code for metal fuel Sodium-cooled Fast Reactors (SFRs) and microreactors. SRT conducts time-dependent radionuclide transport and retention in SFRs for core and ex-core radionuclide source accident sequences. The main objective of SRT is to provide rapid sensitivity and uncertainty analyses, incorporating parametric uncertainties and summarizing probabilistic results. As part of the code validation process, a study focused on the bubble scrubbing module was performed using an experiment recently carried out by the University of Wisconsin-Madison. Based on the analysis, the modeling approach in SRT provides accurate results for small and large aerosols, while slight underprediction of radionuclide aerosol removal are observed for medium sized aerosols. However, the deviation is minor, considering the highly uncertain phenomenon and range of results, and is in the conservative direction. In addition, uncertainty information derived from the experiments is further implemented, reflecting the actual span of parameters, which leads to enhanced agreement with code predictions. The results demonstrate that SRT provides reasonable predictions for the bubble scrubbing process in sodium pool.

The Modular Accident Analysis Program (MAAP), developed by EPRI, allows users to analyze simulated nuclear plant accident scenarios. The code predicts plant responses to severe accidents by evaluating the core, reactor vessel, and containment conditions, and tracks the transport of energy and mass, including water, hydrogen, aerosols, and radioactive species.

The latest version, MAAPv6.00, is being developed in C++ to incorporate modern, state-of-the-art approaches. EDF has contributed to the MAAPv6.00 update to support new designs like Small Modular Reactors (SMRs) that rely on the In-Vessel Strategy for severe accident management. Among EDF updates is the advanced modeling of the corium pool in the Lower Head. This modeling approach includes features such as the kinetics of stratification, which tracks the progressive formation of stratified layers in the pool. This can result in the Focusing Effect, where heat flux concentrates on a narrow lateral surface, potentially exceeding the Critical Heat Flux (CHF) and leading to vessel failure.

Additionally, the distributed modeling of the core support plate enhances heat transfer from the corium pool to the support plate. The sequential melting of the support plate can increase the mass of the metal layer in the corium pool, thereby mitigating the Focusing Effect.

This paper provides a comprehensive description of EDF recent modeling in MAAPv6.00 and presents a use case demonstrating their practical application for severe accident assessments.

The open-source package ‘containmentFOAM’ was developed to provide highly resolved insights, supporting the assessment of the effectiveness of safety measures and possible combustion loads challenging the containment integrity. It comprises a CFD solver and model library developed for and tailored to the expected phenomenology in a large dry PWR containment, as well as tools for input creation and solution monitoring.

This paper aims to summarize the progress made after its first introduction on NURETH-19 (2019). The package was continuously refactored and is currently available as an add-on to OpenFOAM^®-11. Major advancements in the physical modeling capabilities are related to radiative heat transport in participating media, aerosol, and fog transport as well as two-phase flows. Besides, the functional mockup interface (FMI) was implemented, allowing for a flexible integration of system models, packaged as functional mockup units (FMU). Along with the application-oriented validation, best practices were derived and an efficient sensitivity and uncertainty quantification method, based on deterministic sampling, was developed. Concluding, the paper will summarize ongoing applications as well as the strategy for further development.

The evaluation of source terms, which determines the species and quantity of radioactive materials released during a severe accident, is essential for timely safety assessment and the formulation of mitigation strategies. During severe accidents, significant releases of fission products undergo complex integrated phenomena, which inherently introduce substantial uncertainties in the evaluation of source terms. In particular, the behavior of various radionuclides with complex physical and chemical properties significantly increases the uncertainty in the analysis results.

Without quantifying these uncertainties, the reliability of predictions regarding radioactive material release during an accident is compromised, resulting in inaccurate mitigation strategies and safety assessments. Therefore, identifying and quantifying uncertainty factors is fundamental for reliable predictions of source term releases and the development of effective response strategies.

To quantitatively assess these uncertainties, evaluations based on reliable experimental data or actual accident scenarios are required. These experiments offer direct observations of the release behavior of fission products under various accident conditions and provide reference points for accident results.

In this study, the MELCOR code was utilized to benchmark the PHEBUS FPT-1 experiment, assessing core degradation behavior and source term release. Based on these results, key uncertainty variables affecting source term release were identified, and uncertainty analysis was conducted. Additionally, sensitivity analysis was performed to quantify the impact of each variable on the results.

The use of standard deep neural networks (DNNs) have been shown to have better predictive capability than the look-up table (LUT) method to predict Critical Heat Flux values based on input parameters. However, recent work has shown that the model produces unphysical dependence of CHF on the heated length parameter when the heated length parameter is large. We show that this undesired model behaviour is a result of having extremely few data points at high heated length values. One option to resolve this issue is to remove the heated length as an input parameter entirely, but the downside to this is that it removes the possible dependence of CHF on heated length at low heated length values. Consequently, we applied a physics informed neural network (PINN) which penalizes the dependence of CHF on heated length. We scaled this penalty so that it is proportionate to the heated length value. The resultant PINN model had a CHF dependence on heated length only at smaller heated length values and was practically independent of heated length at high heated length values. The PINN model has a lower accuracy on the training data compared to the reference DNN model, which shows that the provided training data strongly implies that there is at least some dependence of CHF on heated length. We studied variants of the penalty term of PINNs and finally obtained a model which had training data accuraries between the LUT method and the reference DNN method.

Despite the tremendous efforts to predict the critical heat flux (CHF), the existing models incorporate remarkable uncertainties due to challenging phenomenological nature and the limited regression feature. An approach applying the artificial intelligence technique for the CHF prediction is expected to overcome the limitations of the conventional methodologies. However, the prediction results by deterministic neural network algorithms, which consist of massive weight/bias matrix in a forms of point values, there are intrinsic concerns in terms of the black-box characteristics, generalization, and reliability for their practical applications. To resolve the concerns inhering in the deterministic approaches, probabilistic neural network frameworks facilitating the quantification of uncertainty and interpretation of their predictions in a wide range of the flow conditions were developed in this study. Three standalone probabilistic neural networks, i.e., Bayesian neural network (BNN), Monte-Carlo dropout (MCD), and Deep ensemble (DE), were constructed to demonstrate the feasibility quantifying the uncertainty information on their CHF prediction. In addition, a series of model-informed neural network architectures, in which the skeptical regression feature in the 2006 CHF look-up table primarily predicts the CHF and neural network models minimize the residual between the actual data and predictions, were developed to improve the generalization capability. The standalone and model-informed deep ensemble frameworks exhibit the best regression and generalization performances providing the aleatoric and epistemic uncertainties in their prediction. Furthermore, influences of the individual parameters and relationships among them are successfully analyzed by application of the interpretable AI technique.

Critical Heat Flux (CHF) is a key design parameter in water-cooled reactors, directly influencing operational safety margins and economic efficiency. However, accurately predicting CHF remains challenging due to its inherent complexity and uncertainty. This study evaluates the performance of two AI/ML models—XGBoost and Multi-Layer Perceptron (MLP—using the NRC CHF database containing approximately 25,000 data points under uniform heating conditions in round tubes.

A robust database splitting methodology was employed to create interpolation and extrapolation datasets for assessing model generalization. Results demonstrated that MLP outperformed XGBoost in interpolation and single-variable extrapolation scenarios. Notably, MLP achieved prediction accuracies comparable to LUT HBM even without explicit training on these data ranges, with improved extrapolation performance driven by feature engineering that transformed the output variable to log(δX). However, MLP exhibited limitations in multi-variable extrapolation regions, with errors approximately three times higher than LUT HBM.

In conclusion, this research demonstrates that AI/ML models, particularly MLPs with optimized input-output features can serve as robust alternatives to traditional LUT methods for CHF prediction in round tube geometries. Future work will address multi-variable extrapolation challenges and extend these methodologies to more complex geometries like rod bundles for broader applicability in reactor safety analysis.

Deep generative models (DGMs) are powerful deep learning models for generating synthetic but realistic data by learning the underlying distribution of a training dataset. DGMs offer a potential solution to the challenges of data scarcity and data imbalance, which are very common in nuclear engineering as the measurement data is often obtained from costly experiments. Diffusion models (DMs), a relatively new family of DGMs, have demonstrated great potential in data augmentation especially for images and videos. In this work, we explored the effectiveness of DMs in generating scientific data for nuclear engineering applications. Our focus is on evaluating the performance of DMs in generating critical heat flux (CHF) data, using a training dataset that was originally used to develop the 2006 Groeneveld lookup table. The DM is assessed on its ability to capture the correlations between different parameters in the dataset and whether it generates physically meaningful values for each parameter. Additionally, we compared the full joint empirical cumulative distribution functions (ECDFs) of the real and synthetic datasets to evaluate the overall distributional similarity. The results show that DMs successfully generate CHF data by accurately learning the correlations between parameters without producing unphysical samples. The ECDF comparison further confirms that the synthetic data closely matches the measurement data, demonstrating the potential of DMs for data augmentation in nuclear engineering.

The development of small modular reactors responds to a need for more flexible energy generation for a wider range of applications. Within the ELSMOR European project, the integrated E-SMR light-water-reactor concept consisting in a tank containing the core, pressurizer and compact steam generators has been investigated. The tank is in a metal enclosure, itself immersed in a large pool. The aim of this work is to study the pool thermal-hydraulic behavior by means of the 3D module of CATHARE-3.

Within the first phase of calculations, a standalone analysis of the water pool is carried out using the CATHARE-3 3D module. The containment is represented by a fixed temperature boundary condition. It is found that the evolution of the liquid temperature distribution is uniform across the pool as long as the water is in contact with the containment wall. A further refinement of the pool nodalisation does not significantly improve the results.

Within the second phase of calculations, the previous 3D water pool model is coupled with the containment, the mass and energy releases being taken from ASTEC and MAAP severe accident calculations. Two accidental sequences are considered: the first one involves the evacuation of the decay heat through the containment walls, while the second involves the use of a passive condenser. The results obtained by CATHARE are comparable to the tendencies predicted by ASTEC and MAAP. Concerning the 3D pool behavior, a uniform liquid temperature distribution is observed in the first accident, while the second one shows a temperature stratification.

High Power Density Pressurized Water Reactors (HP-PWRs) offer significant advantages in terms of thermal output within compact volumes, making them a promising option for applications such as small modular reactors. However, under high-power operating conditions, the occurrence of high local void fractions within HP-PWR cores presents unique challenges, affecting both the neutronic and thermohydraulic behaviors. This paper introduces a high-fidelity, fine-mesh neutronics-thermohydraulics (N-TH) coupling method to address these challenges for accurately modeling core behavior under high local void fraction conditions. The method incorporates flow distribution calculations (FDC), which significantly improve simulation accuracy by overcoming the limitations of traditional methods that assume uniform flow distribution. Our results show that under two-phase flow conditions, the introduction of FDC significantly alters the void fraction distribution, as well as the fuel and cladding temperatures, compared to traditional methods. Specifically, under 100% full power conditions, the power level of the hottest assembly decreased by approximately 0.8%, the mass flow rate of the hottest channel decreased by 12.87%, the maximum fuel temperature dropped by 0.77 K, and the maximum void fraction increased by 0.144. The impact of FDC is more pronounced in two-phase conditions and minimal under single-phase conditions. This study provides a valuable tool for the design and optimization of HP-PWRs and offers insights into improving reactor power density.

In U-tube steam generators employed in pressurized water reactors, flow-induced vibrations within the upper cross-flow U-section of the bundle are a major cause of fatigue and equipment damage. As it evaporates flowing upward and the steam quality increases, the water-steam mixture on the shell side transitions from bubbly flow to the intermittent and annular flow regimes. The local regime significantly influences the force exerted on the tubes. Consequently, accurate knowledge of the local flow conditions is crucial for assessing flow-induced vibration, but the consistent numerical prediction of the two-phase flow regime remains a major challenge for available CFD methodologies.

In this paper, the morphology-adaptive GEMMA (GEneralized Multifluid Modelling Approach) CFD model is used to predict the flow across a horizontal 7 x 5 cross-flow tube bundle in the bubbly and intermittent regimes. The GEMMA model, implemented in OpenFOAM, is based on the multifluid framework, but partially resolves interfaces over a certain length scale. This enables GEMMA to handle the entire spectrum of interface length scales in all flow regimes, which is traditionally challenging for available CFD approaches. In the intermittent regime, unsteady large gas structures are successfully predicted, and this enables a more accurate estimation of the void fraction inside the bundle. The intermittent nature of the local flow is reflected in the predicted force exerted on the tubes. The use of the GEMMA model results in much more time-fluctuating forces on the tubes, compared to the standard dispersed phase two-fluid model, unable to predict the intermittency of the flow.

The growing integration of renewable energy into electricity markets is driving nuclear power plants to shift from traditional baseload operation to more flexible modes. Flexible operation of nuclear reactors necessitates the evaluation of several technical challenges, including Xe-135 oscillations, a fission product that can significantly impact the reactor's operational stability.

This study focuses on analyzing Xe-135 oscillations triggered by load variations during the flexible operation of nuclear reactors. Flexible operating conditions are implemented in a 3D thermohydraulic-neutronic model of a PWR-KWU reactor core using the coupled code RELAP5/PARCSv3.2. Key parameters, such as the Axial Offset (AO), are examined to assess spatial distortions in power and Xe-135 distribution within the reactor core.

The results of this study highlight how variations in Xe-135 concentration affect the process of load increases and decreases during the flexible operation of PWR-KWU reactors.

The IAEA CRP on “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop” aims to support and achieve validation of computational fluid dynamics (CFD), subchannel, and system thermal-hydraulics (STH) analysis codes for heavy liquid metal systems. In particular, the Benchmark consists of two reference cases used for model training and a blind case to be reproduced for sake of model validation and accuracy assessment.

Together with stand-alone codes applications, a whole work package is devoted to the analysis of the addressed scenarios adopting multi-scale STH/CFD coupled applications. The CRP participants addressed the common problems adopting different system thermal hydraulics code and CFD codes, also considering different assumptions regarding the boundary conditions and involved phenomena representation. In particular the CFD approach was adopted for the simulation of the Fuel Pin Simulator, which represents a key component of the NACIE-UP loop, while the STH was considered for the remaining sections of the facility. The use of CFD for the FPS should allow for a better representation of the involved heat transfer and friction phenomena as well as the capability to obtain refined predictions of local wall and bulk temperature distributions in transient conditions. On the other hand, the STH approach allows for a relatively small computational cost of the other facility components.

The present paper reports on the results obtained by the CRP participants providing comments and improvement suggestions for the liquid metal loop modelling.

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.

System-scale TH (STH) analysis codes have extensively been used in WCNR safety evaluation along with quantifying the prediction uncertainties in close conjunction with adopting the best estimate (BE) safety analysis. There still exist some deficiencies in the BE safety evaluation, however, originating mainly from our limited knowledge or poor understanding of underlying fundamental physics on key TH phenomena associated with two-phase flow hydrodynamics and heat transfer, which are broadly relevant to WCNR safety concerns.

TH experiments and analyses for WCNR performance analysis and safety evaluation, in general, need to be carefully checked in terms of their scalability to assure whether they are realistically representative of prototypic situations. The basic concern of ‘scalability’ originates from the differences or gap existing between the prototypic and down-scaled systems due to their idealization and/or simplification.

The scalability of experimental data used for validating TH analysis codes will be discussed, focusing on STH codes with their application to WCNR safety evaluation accompanied by the uncertainty quantification.

Discussion will be focused mainly on our unsatisfactory understanding of fundamental physics associated with the constitutive relations adopted in STH codes, many of which were developed based on unrealistic observation under non-prototypic geometric and TH conditions, and partly on the limited numerical capabilities of STH codes in describing multi-dimensional features of dominant phenomena. Then the perspectives of advanced TH safety evaluation are introduced aiming at improving the modelling and simulation (M&S) capabilities.

Results of a Best-Estimate Plus Uncertainty analysis of the LOFT L2-5 experiment performed with millions of cases is presented. The results are used to examine how the techniques traditionally used in analyses are equipped to handle and address likelihoods significantly less probable than at the 95%/95% level. The paper describes the required changes to the underlying probability distribution functions that were required to ensure physical results. The paper presents changes to the model required to achieve sufficient robustness for the process and changes to the typically used uncertainty distributions.
The analyses show that the simulation results do not form a single population of data. The input distributions that lead to the various clouds of data are presented. Physical explanations are provided that link the specific changes in uncertainty parameters to the various outcomes.

Best Estimate Plus Uncertainties (BEPU) approaches are often perceived as complex in licensing processes by licensees. The complexity of the BEPU approach is generally considered justified, as it a priori offers the potential for a more accurate estimation of safety margins. Since it tends to be less conservative than deterministic methods, it raises legitimate questions about its maturity from a regulatory standpoint, particularly given the challenges of nuclear safety assessments.

A relevant example is the case of Loss of Coolant Accidents (LOCA), where BEPU methodology can play a crucial role in assessing fuel behavior (e.g., peak cladding temperature, rupture…). Analyzing these transients involves multiscale (from sub-channel to reactor level) and multiphysics phenomena (multiphase thermohydraulics, fuel and cladding thermomechanics, neutronics, etc.).

Recently, BEPU methodologies have been proposed by licensees in France for the safety demonstration of operating and newly designed reactors. IRSN’s analyses have led to two key questions:

• Can numerical modeling be considered "best estimate" given the diversity of phenomena involved?
• Can we fully rely on the estimation of uncertainties?

This paper highlights some of IRSN’s concerns regarding these aspects, drawing on expert judgments or explicit CATHARE modeling. IRSN believes that BEPU methodologies could play a role in safety demonstration due to their ability to naturally incorporate different combinations of multiscale and multiphysics phenomena. Nevertheless, using BEPU does not exclude some penalties to be required for covering certain limitations.

Since the 1980s, the Best Estimate Plus Uncertainty (BEPU) methodology has been a cornerstone for deterministic safety analysis of design basis accidents in nuclear power plants. Despite endorsements from the International Atomic Energy Agency (IAEA) and various national regulatory bodies, its industrial application remains limited. At the BEPU 2024 conference in Lucca, Italy, a panel of global experts convened to discuss strategies for expanding BEPU’s industrial use. The panel session focused on:

• Enhancing understanding of uncertainty analysis and Input Uncertainty Quantification (IUQ).
• Addressing challenges and exploring future advancements in BEPU methodology.

• Sharing insights and experiences in BEPU application.

• Fostering dialogue on extending BEPU to broader industrial applications.

Key outcomes of the discussions included:

• Advancements in UQ/IUQ methods by NEA, highlighting the need for international cooperation.
• The role of IUQ in boosting confidence in BEPU.
• Evolution of BEPU methodologies for LBLOCA, addressing high burnup fuels and ECCS rule revisions.
• Expansion of BEPU applications from LOCA safety analyses to non-LOCA scenarios, severe accidents, containment integrity, and dose consequences.
• The necessity of a graded approach for broader BEPU application.
• Limited BEPU use in some countries due to regulatory conservatism and high development costs.
• Potential benefits for non-LWR applications, necessitating new experimental data.

This paper summarizes the main contents and outcomes of the panel discussions.

In order to solve the problem that the planned life extension of Qinshan No.3 Nuclear Power Co., Ltd. in China (hereinafter referred to as Qinshan No.3 Nuclear Power Plant) leads to the increase of spent fuel, and the capacity of existing spent fuel dry storage modules is insufficient, based on the original 1~6 (QM-400) spent fuel storage modules, the intensive spent fuel dry storage facilities (M1 and M2 spent fuel storage modules) have been developed. Compared with QM-400 spent fuel storage module, M1 and M2 modules have larger storage capacity and higher energy density. In order to demonstrate the thermal safety of M1 and M2 modules, a thermalhydraulic program is used to establish the thermal analysis model of M1 and M2 modules based on conservative initial assumptions, and calculate the temperature of each region under normal operation and accident analysis of the module under extreme weather conditions. At the same time, the three-dimensional fluid CFD program is used to verify the calculation results of the thermalhydraulic program, and the calculation results of thermalhydraulic program and CFD program are integrated, The thermal safety of M1 and M2 modules is demonstrated. Finally, from the perspective of engineering feasibility and the condition of nuclear power plant site, M1 module is adopted as the implementation plan of intensive spent fuel dry storage facility.

There is currently a growing interest in analysing the behaviour of Advanced Technology Fuels (ATF) under development. Among the new evolutionary ATF designs, the FeCrAl and Cr coated claddings are the most promising. On the other hand, the LBLOCA sequences in Pressurized Water Reactors (PWR) are among the most demanding for safety systems and have a small safety margin. To perform this analysis, NFQ and UPM developed an in-house version of the TRACE5P6 system code for FeCrAl cladding, as TRACE5P6 is not designed to simulate this cladding material. Then, a BEPU analysis of LBLOCA sequences in a PWR was performed for Zry, FeCrAl and Cr-coated Zry cladding and the safety margins were obtained for each case. The results show that the safety margins for ATF materials are greater than those for the Zry case.

The steam system downstream of the main steam isolation valves (MSIV) is normally not structurally verified for the hydraulic loads that can occur following a steam line break (SLB). Also, the turbine trip is normally not classified according to nuclear safety grade standards. Mechanical failure of the steam system, or a failure of the turbine trip system, can therefore not be excluded following a SLB. This could lead to additional steam outflow in addition to the break flow. As a consequence, blowdown of two steam generators could occur if a single failure is assumed on one MSIV. Also, considering extreme external events such as an earthquake or antagonistic actions, the integrity of the turbine building itself, along with the whole steam system outside containment, could be questioned, potentially leading to blowdown of all three steam generators.

In NUREG-0138 it is expected that the assumption of a stuck control rod would compensate for any penalties associated with the blowdown of two steam generators. In the present study, this statement has been investigated using the system thermal hydraulics code TRACE with its built in neutronic point kinetics model. SLBs with blowdown of one, two and three steam generators are analyzed. The effect of the stuck control rod assumption is also studied. An increase in maximum power is observed when blowdown of several steam generators is assumed. However, as expected in NUREG-0138, a large decrease in maximum power is seen if no stuck rod is postulated. Finally, the impact on DNBR is discussed.

The Japanese fusion reactor, JA-DEMO, is designed to generate electricity at approximately 300 MW. The amount of enthalpy stored in the reactor's coolant will be significantly larger than that of ITER's. Consequently, we must consider the potential risk of a Loss of Coolant Accident (LOCA) in the JA-DEMO reactor. In this research, we conducted a thermohydraulic LOCA analysis of the Japanese water-cooled DEMO reactor, JA-DEMO, with TRACE code. The U.S. NRC has developed a TRACE code for LOCA analysis of light water reactors. We analyzed three distinct LOCA scenarios: In-Vessel LOCA, Divertor LOCA, and Ex-Vessel LOCA. In the In-Vessel and Divertor LOCA scenarios, water enters the plasma chamber (PC) from the outer blanket and divertor, respectively. In the Ex-Vessel LOCA scenario, water enters the vault from a pipe in the primary cooling system. Initially, we conducted a conservative analysis assuming the maximum break area in each coolant pipe, primarily observing pressure transients in the plasma chamber and vault. Subsequently, we tuned several parameters like pipe break areas for parameter surveys, finding that the break area must be smaller than a threshold in In-Vessel LOCA to maintain the PC pressure below the design pressure of 0.5 MPa. Additionally, we explored optimal component geometries to minimize the impact of LOCA in JA-DEMO.

Magnetohydrodynamics (MHD) investigates the behaviour of conducting fluids interacting with magnetic fields, such as the liquid metals foreseen in the blanket of many fusion reactor designs. The intricated physics involved in MHD scenarios often results in significant computational costs. In this regard, Reduced Order Modelling (ROM) methods may represent a promising solution, as they can approximate complex systems with lower-dimensional yet still-accurate models especially in multi-query and real-time contexts. One of the most famous techniques is the Dynamic Mode Decomposition (DMD), a data-driven algorithm designed to learn the best linear model based on time series datasets. In this work a parametric version is applied, which treats DMD operators as snapshot data, mapping parameter values to modal coefficients. This framework allows for the efficient capture of transient dynamics across a range of parameters, improving computational efficiency and accuracy. This approach is applied to a MHD scenario involving compressible lead-lithium flowing in a channel subjected to different magnetic field intensities, which represent the varying parameter. The channel includes regions on the walls at different temperatures to investigate the effects of various magnetic configurations on the thermo-hydraulics of the liquid metal. This study represents an application of a promising ROM technique to an advanced thermohydraulic scenario, involving conductive fluids influenced by magnetic fields. The results show that the parametric DMD significantly reduces the computational burden while keeping a desired accuracy in predicting the complex MHD flows, highlighting its potential for broader applications in fusion technology and MHD systems.

Inductionless magneto-convection is directly simulated in a Rayleigh-Bénard configuration using the high-fidelity finite-difference solver ‘Xcompact3d’. The Rayleigh numbers considered are in the range whilst the Hartmann numbers (Ha) are in the range 0-1000. Two Prandtl numbers are considered; Pr=0.71 corresponds to air whilst Pr=0.025 corresponds to the liquid LiPb eutectic present in fusion breeder blanket systems

In the presence of no magnetic field the flow is turbulent, chaotic and unsteady. Applying a magnetic field leads to a dramatic reduction in turbulence levels, with steady laminar flow observed in the high Ha limit. Field orientation is a critical factor; wall-parallel fields lead to ‘quasi-2D’ turbulence where there is little variation in the flow along the field direction, whilst wall-normal fields lead to 3D structures that are significantly damped along all three of the spatial directions. In the wall-normal case a significant degradation of the heat transfer performance is observed with increasing Ha whilst the wall-parallel field has little influence on the overall heat transfer performance.

A physics focused analysis is then conducted with a focus on the coherent turbulent structures present in the system, and in particular how the strength of the applied magnetic field influences these structures. The analysis is conducted using conditional averaging to understand the separate roles of ejecting and impacting plumes through the perspective of turbulence budgets. Additionally, a spectral analysis is conducted to understand the most dominant structures in the flow and the roles of these structures in the recurring cycle of magneto-convection.

Achieving fusion reactor design goals require transient analyses to assess conditions during operational and accident scenarios. Analysis of cyclic behavior is essential for plant operation considerations and component lifetime predictions. While transient analysis requires consideration of start-up, shut-down and disruptions, pulsed operation also involves oscillating reactor power. The frequency of oscillations determines thermal inertia and stresses acting on the materials and the blanket. Engineering analysis must therefore include time-dependent loads. Simulating transients are traditionally computationally prohibitive, especially for fluid flow and heat transfer analysis. A major challenge and bottleneck for high fidelity pulsed simulation is turnaround time (currently months for a single ITER discharge). In this work we develop a flexible framework and utilize exascale computing to enable high-fidelity transient simulations.

Accurate modeling of the heat source in the reactor to ensure safe operations is done using tools (OpenMC, MCNP) developed through the FERMI project. While the power level of the fusion reactor determines the magnitude of the resulting neutronic heat deposition, the irradiated structural materials generate decay heat during and after the pulse. These are calculated using the aforementioned tools. Both pulsed and steady state concepts require active cooling during operation, maintenance, and shut-down. Transient thermal hydraulics analysis of the various components is performed to include the decay heat obtained from neutronics. These calculations require frequent data transfer of the volumetric heat deposition from neutronics to the conjugate heat transfer module. Sensitivity of the data exchange frequency will be studied to assess the optimum rate without loss of accuracy.

The Breeding Blanket (BB) is a fundamental component for fusion reactors, responsible power production, neutron shielding, and tritium generation for sustaining fusion. For DEMO, the Water Cooled Lithium Lead (WCLL) and Helium Cooled Pebble Bed (HCPB) designs are leading candidates. ITER plays a pivotal role in validating these BB concepts, using Test Blanket Modules (TBMs) to evaluate their functionality under reactor conditions, providing data on performance, efficiency, and safety.

To provide a validation for WCLL BB concept components, as well as characterization of mock-ups and portions of the BB on a relevant scale, Water Loop (WL) facility is currently under development at the ENEA R.C. Brasimone. The WL facility comprises three thermally coupled loops. The first loop emulates the DEMO Primary Heat Transfer System (PHTS) thermal-hydraulic conditions, operating with water ranging between 295-328°C at 15.5 MPa. This loop is featured with flanges enabling the non-simultaneous connection with different Test Sections (TSs), thereby enhancing its versatility. The TSs can be tested in different operative conditions including inside a Vacuum Chamber (VC), where they undergo irradiation by an electron beam gun aimed at replicating fusion reactor heat flux and simulating the tokamak environment or in connection with a PbLi loop, in order to simulate normal or accidental conditions. The secondary and tertiary loops are primarily tasked with dissipating this power, ultimately exchanging it with a cooling tower.

The present presents a comprehensive overview of the facility layout and requirements, and a RELAP5/Mod3.3 characterization of the facility main thermal-hydraulic parameters under operative conditions.

Tokamaks are inherently pulsed fusion reactors due to the transformer function of the central solenoid inducing plasma current. Non-inductive current methods try to provide steady plasma current, however the efficiency is currently an open issue and therefore their benefit could not compensate the detrimental cost when pursuing steady-state power operation with tokamaks.

An alternative is the so-called Indirect Coupling Design of the Balance of Plant System where the plasma pulsed operation is decoupled from the Power Conversion System of the Fusion Power Plant by using an Intermediate Heat Transfer System (IHTS) hosting an Energy Storage System (ESS). This is actually the reference option selected for the He-cooled EU-DEMO Design.

Presently the HELOKA-Upgrade Storage experimental project is in construction at the Karlsruhe Institute of Technology aiming at studying the behavior of such indirect concept in a mock-up facility. The functionality and operability of the IHTS during normal EU-DEMO operation will be investigated. The first phase of the project consists of a molten salt (MS) loop with an ESS coupled to a water-cooling system acting as heat sink, where the MS loop heat source is an electrical heater. Heat transfer measurements will be performed in a test section undergoing similar conditions as in the Helium-MS Heat Exchanger of EU-DEMO Design. In a later phase, the heat source will be the existing high-temperature Helium loop.

The present paper presents pre-test analysis performed with SIM-code to assess the thermal-hydraulic behavior of the MS and water loops supporting the experimental campaign.

The operation of the ITER reactor will represent a milestone in nuclear fusion research, serving as crucial step towards the realization of commercial fusion energy production by bridging the gap between current research efforts and future industrial-scale deployment. A key component of a fusion reactor is the Breeding Blanket (BB) that must generate tritium fuel, shield the vacuum vessel from high-energy neutrons and transfer the heat generated by the plasma to the power conversion system. One of the proposed BB concepts is the Water-Cooled Lithium Lead (WCLL) which is going to be tested under realistic fusion reactor conditions in ITER in the form of a Test Blanket Module (TBM).

In this framework, at the ENEA R.C. Brasimone the construction of W-HYDRA, an experimental infrastructure dedicated to the investigation of the water and lithium-lead technologies is ongoing. As part of W-HYDRA, Water Loop (WL) facility will investigate the WCLL technology and thus a 1:1 scale mock-up of the WCLL BB TBM will be hosted and experimentally studied. The design characteristics and performance of the TBM will be assessed to provide valuable experimental results in view of ITER operation.

The present paper is focused on the thermal-hydraulic numerical study of the TBM component within WL using RELAP5/Mod3.3. Specifically, the study investigates Normal Operation State conditions (i.e., pulse-dwell and dwell-pulse transients) and accidental scenarios (i.e., Loss Of Feedwater Accident, LOFA), aiming to provide preliminary insights into the operation of WL and investigate the control strategy for conducting the TBM experimental campaigns.