Conference Agenda

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 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.

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.

The validation of decay heat removal (DHR) systems and the characterization of thermal hydraulic phenomena in the plena of the liquid metal-cooled, pool-type research reactor MYRRHA, under design at SCK CEN, the Belgian Nuclear Research Centre, is accomplished by experiments and numerical investigations. For this purpose, the E-SCAPE facility at SCK CEN is a thermal hydraulic 1/6-scale 3-D model of the primary system of MYRRHA, with an electrical core simulator, and cooled with Lead Bismuth Eutectic (LBE). Its scaling is based on the preservation of the overall system behavior and the reproduction of the major thermal hydraulic phenomena under DHR conditions. Results from steady-state and transient thermal hydraulic experiments in forced and natural circulation demonstrate that in loss of flow conditions, a natural circulation flow establishes, driven by buoyancy, that can remove the core power, with fuel cladding and reactor structure temperatures that are within safety limits. In natural circulation, thermal stratification occurs in the upper plenum of the facility. System Thermal Hydraulics (STH), Computational Fluid Dynamics (CFD) and coupled STH/CFD models of E-SCAPE have been built in different phases of its lifecycle for scaling, design, pre-test and post-test analyses. The results of post-test simulations are compared with experimental data, showing that the phenomena driving the flow are accurately represented in both the numerical models and the experimental facility. This paves the way for the validation of numerical tools used in the safety analyses of MYRRHA

The complexity in nuclear reactor safety issues has highlighted the need for more flexible and robust modeling approaches. Particle-based methods—such as Smoothed Particle Hydrodynamics (SPH), Discrete Element Method (DEM), and Lagrangian Dispersion Model (LDM)—offer significant advantages in modeling highly nonlinear, multiphase, and multiscale phenomena that challenge conventional grid-based methods. This paper presents the basic principles of these particle-based techniques and discusses their implementation within high-performance computing (HPC) environments, with an emphasis on graphical processing units (GPU)-based parallelization strategies. The capabilities of particle-based frameworks are demonstrated through a series of nuclear safety applications, including in-vessel retention and external reactor vessel cooling (IVR-ERVC), corium spreading, core catcher impact analysis, steam explosions, and environmental radionuclide dispersion. These case studies illustrate the methods' potential to handle complex interfaces, large deformations, and strongly coupled multiphysics interactions without explicit interface tracking. The paper concludes by outlining current limitations—such as computational cost, turbulence modeling, and phase-change physics—and suggests future directions toward establishing particle-based approaches as integral tools for next-generation nuclear safety analysis.

Molten salt reactors (MSRs) are a promising reactor type, offering excellent safety and economic benefits due to the stable properties of molten salt coolant at high temperatures. However, the relatively high freezing point of molten inorganic salts poses a risk of coolant solidification, potentially blocking pipelines when flowing through colder sections. This study investigates the process of molten salt filling in cold pipes, including solidification-melting behaviors, and analyzes the pressure drop variation to estimate pipe clogging caused by freezing. Using the Volume of Fluid (VOF) method and the enthalpy-porosity model, the commercial CFD code ANSYS Fluent is employed to numerically simulate the filling process. Results reveal that a solidification layer forms near the cold wall, while the high-temperature incoming flow melts the layer first at the inlet, with the layer thickness decreasing along the flow direction. Analysis shows that the mushy zone constant (A_mush) significantly impacts flow pressure loss, particularly for lower inlet temperatures in finite-length pipes. Higher values of ​ accelerate pressure loss growth, though this increase remains below the maximum upstream pressure head. Comparison with experimental data from Zhang W. demonstrates that the estimated blockage penetration distances for HTS at A_mush=5×10⁴ exhibit an error within 30%. This highlights the critical importance of selecting an appropriate mushy zone constant to accurately predict solidification processes when using the enthalpy-porosity method.

In molten salt reactors (MSRs), liquid fuel offers benefits like high economic efficiency, safety, and low radioactive waste. This fuel, typically a fluoride- or chloride-based salt, contains soluble fissile material in a carrier salt. Compared to water coolant, the working fluids in MSRs have higher melting points and greater corrosivity. Insoluble fission products generated in the core interact with these fluids, which can threaten the integrity of reactor structures such as pumps. Simplifying the primary system is proposed to enhance MSR safety and integrity. This study introduces the passive molten salt fast reactor (PMFR) to simplify the primary system. The PMFR design removes pumps and relies on natural circulation, increasing safety and simplifying reactor design. However, overly simplified core designs can cause imbalanced flow and unexpected heat removal, affecting reactor power. Therefore, stabilizing core flow while minimizing pressure drop is essential. This paper validates the guide structure performance of PMFR using the multi-physics code GeN-Foam, based on OpenFOAM, which models various physics, including neutronics, thermal-hydraulics, and structural thermal-mechanics. Long-term pseudo-steady operation simulations of PMFR demonstrate its feasibility in achieving target power. Results show encouraging performance under normal operating conditions and suggest further improvements to enhance PMFR safety and economics.

The international maritime organization (IMO) has imposed serious restriction on the carbon emission of the commercial shipping industry, which now accounts for nearly 5% of the world amount. Nuclear power can be an important alternative for supplying the large ships with long-durance and near-zero-carbon-emission energy. The Fluoride-salt-cooled High-temperature Reactors adopting the low-operation-pressure fluoride salt as the coolant and the TRISO-particle fuel, show great advantages in the inherent safety, high economics, and reduced difficulty in the licensing. Thus the FHRs can be good reactor concept candidate for the commercial ships. The wind and wave in the ocean bring about oscillation to the flow in the reactor system, which leads to the periodic variation of the heat transfer between the salt and fuel. In further, the safety performance of the shipping-applied FHRs will also be influenced. The Nuclear Reactor Thermal-hydraulic Lab in Shanghai Jiao Tong University has explored the influence of the ocean environment on FHRs, including the core thermal-hydraulics and safety evaluation. The flow regime transition under the pulsation flow is explored, with the transition criteria determined for different pulsation amplitude and period. The system analysis code is developed with the implementation of the additional force models. The system code is validated through the scaled integral effects experimental data. Finally the safety performance under the inclination, heaving and rolling motions is obtained.

In light of the licensing process of advanced reactor designs, a fundamental step to support the safety assessment consists of identifying and quantifying the uncertainties resulting from a lack of extensive practical knowledge and modelling assumptions. The uncertainty characterisation imposes specific requirements for the numerical tools employed to inspect safety-related phenomena. When dealing with Molten Salt Reactors (MSRs), the inherent characteristics of circulating fuel result in the need to perform multidimensional and multiphysics simulations to investigate the steady state and dynamic behaviour of the MSR concept. The multiphysics approach allows to capture the relevant governing phenomena strictly related to the strong coupling between neutronics and thermal-hydraulics. On the other hand, in the context of safety analysis, system codes have proven their suitability to represent the whole plant behaviour, implement submodules devoted to uncertainty quantification, and calibrate models with experimental data. In this work, a computational chain coupling well elaborated system codes and high-fidelity multiphysics tools is developed to manage in the same environment different levels of detail. The modelling framework couples Modelica and OpenFOAM modelling tools thanks to Functional Mock-Up Interfaces, which define a container and an interface to exchange dynamic simulation models. This approach embraces a multidimensional and multiphysics model of the MSR core while preserving a global representation of the plant. The OpenFOAM-Modelica coupling chain is tested on a case study involving a symmetric portion of the Molten Salt Fast Reactor primary loop with a simplified representation of the intermediate salt circuit and Balance of Plant.

The Molten Salt Reactor (MSR) represents a prominent Generation IV design, addressing the urgent need for safer and more sustainable nuclear energy production. This study aims to capture the multi-physics behavior of the Molten Salt Reactor Experiment (MSRE), with a particular focus on thermal-hydraulic and neutronic interactions within the primary loop. Utilizing the GeN-Foam code, we implement detailed Computational Fluid Dynamics (CFD) and heat transfer models to enhance the accuracy of turbulence, drag forces, and porous media characteristics.

Benchmarking against data from Oak Ridge National Laboratory (ORNL) confirms the robustness of this approach, with simulation values closely aligning with recorded ORNL data. For instance, the fuel velocity in the core exhibited a deviation of merely 0.84% from ORNL data, while the pressure at the MSRE core was within 0.964% of the recorded values. Furthermore, temperature measurements at the fuel inlet and outlet demonstrated minimal deviations of 0.039% and 0.008%, respectively. These results provide critical safety insights by elucidating feedback mechanisms that influence neutronics, thermal-hydraulics, and structural integrity.

Significantly, this model, based on the established multi-physics framework of GeN-Foam, can be adapted for other MSR designs through modifications to geometry and input parameters, obviating the need for further code development. The findings from this research offer valuable insights for optimizing MSR designs and safety evaluations, thereby contributing to future regulatory and developmental applications in the field of nuclear technology.

This paper describes implementation of Molten Salt – Concrete Interactions (MSCI) into the System Thermal-Hydraulic code SPECTRA. It consists of two parts:

• The first part discusses the Molten Core – Concrete Interactions (MCCI) model, applicable to the LWR fuel, which had been implemented at an earlier stage.
• The second part discusses the MSCI model, applicable to MSRs.

The summary and conclusions are presented below.

• MCCI. The MCCI model appropriate for LWR fuel was validated based on comparisons with other code (MELCOR) results. Recommendations for future work include:
  • Implementation of reactions such as melting of quartz, decomposition of hematite into magnetite, melting of magnetite.
  • Implementation of reduction of SiO₂ and Fe₂O₃ by Zr.
  • Validation against available experimental data.
• MSCI. Due to lack of suitable data, a validation of the model by comparison with measured data or even code-to-code comparisons is currently not possible. Two test runs were performed as a verification and qualitative validation of the model.
  • Verification is performed by comparing MSCI results with MCCI results in a hypothetical case where the fuel-salt mixture properties are the same as the corium (UO₂-ZrO₂) properties.
  • Qualitative verification is performed by using realistic (as far as possible) data and checking if the qualitative behavior is in agreement with what is expected.

The safety performance of SFR designs is commonly assessed through Unprotected Loss of Flow (ULOF) transients where active core cooling systems are lost. The importance of ULOF transient relies in its potential to progress into the coolant boiling phase and eventually into partial/even total core destruction. It requires the detailed consideration of the particular effects of various specific design characteristics (e.g., upper sodium plenum, absorber layers, discharge tubes, etc.) during the progression of the transient under consideration.

This work presents the results of SAS-SFR simulation for a 10 s halving time ULOF transient including transient power, reactivity effects and fuel thermal-mechanical and coolant thermal-hydraulic conditions. The SAS-SFR model used provides a precise description of the accident progression in all SA-channels, thus results of the first failing SA-channel are presented in detail to give a deeper insight of the physical phenomena taking place during the various accidental phases.

SAS-SFR calculations predict boiling onset in all SA-channels followed by clad motion in 30 out of 34 channels and fuel pin break-up in 20 out of 34 channels by the end of the calculation. Clad relocation does not block the coolant channel completely, thus after fuel break-up, mobile fuel is relocated outside the core and a strong negative reactivity shuts down the reactor without damaging the hexcan. Therefore, SA hexcan integrity is assured although pin failure cannot be avoided in the SFR low void core design analyzed. However unless cooling conditions are improved reestablishing the coolant flow, cladding integrity will be at risk.

Predicting tritium transport and permeation is a critical challenge in Fluoride Salt-cooled High-Temperature Reactors (FHRs). The coolant, Flibe (2LiF-BeF₂), generates significant tritium via neutron transmutation, which can permeate into the secondary system through heat exchangers. Accurately estimating the multi-dimensional tritium permeation rate is essential, particularly given the increasingly complex geometries of heat exchangers, necessitating a robust numerical tool. To address such technical needs, we have developed scalarMultiRegionFoam, a novel three-dimensional OpenFOAM solver. Built upon chtMultiRegionFoam, the solver incorporates scalar transport equation for the fluid domain and diffusion equation for the solid domain. Additionally, we have implemented a new boundary condition, scalarTransportCoupledMixed, to accurately model the fluid-solid interface. We validated the developed solver against three cases. First, we verified the governing equation within a single domain by comparing it against an analytical solution for a transient problem. Next, we assessed the boundary condition for the fluid-solid interface using a steady-state analytic solution. Finally, we validated the complete model against transient experimental data. The verification and validation results have demonstrated the reliability and accuracy of our solver, establishing it as a powerful tool for simulating 3D tritium transport in FHRs.

Reliability and stability are crucial for nuclear reactor operations, especially in a small-scale system such as micro-reactors, where instability can lead to power fluctuations resulting in localized overheating. Heat pipes often encounter instability during transient phases such as startup and load-following operations, which can induce geyser boiling. This phenomenon occurs when subcooled liquid is expelled to the condenser section, resulting in significant temperature oscillations. Understanding the instability mechanism in heat pipes is necessary to optimize heat pipe micro-reactor operation. This study presents a transient thermal performance of a closed water thermosyphon – a wickless heat pipe, serving as a preliminary step towards modeling a heat pipe for micro-reactor applications. The simulations are conducted using OpenFOAM, a Computational Fluid Dynamics (CFD) code with a Volume of Fluid (VoF) solvers to capture the two-phase flow and phase changes in a thermosyphon. This research evaluates various initial filling ratios, working fluid temperatures, and heating powers in a 2D domain to identify optimal conditions for mitigating thermal instabilities during the startup phase. The OpenFOAM model was validated by comparing it with existing literature that utilized ANSYS Fluent, and it gave us similar results. The CFD modelling and the results of this study will contribute to a better understanding of the thermosyphon’s behavior in transient conditions, serving as the preliminary study for future investigations into more complex heat pipe micro-reactor modelling. In addition to that, this research uses OpenFOAM, which is uncommon in existing literature, which will address a gap in existing literature.

In recent years, significant research and development efforts have focused on various aspects of supercritical water-cooled reactors (SCWRs), with a particular emphasis on thermal-hydraulic analysis. Computational fluid dynamics (CFD) modeling has been extensively used to predict the thermal-hydraulic behavior within SCWR fuel assemblies. This modeling is crucial for validating heat transfer characteristics near critical and pseudocritical points, especially when operating at pressures below the critical threshold. One key focus of this research is simulating trans-critical transients, where the system pressure drops from supercritical to subcritical levels, to understand the effects on the fuel assembly.

To ensure the accuracy and reliability of these CFD models, experimental data from simpler geometries such as single tube and small rod bundles are often used for validation. The available experimental setups provide valuable insights into the behavior of heat transfer under supercritical and subcritical conditions, enabling better predictions and optimizations for more complex fuel assembly designs. By leveraging both CFD modeling and experimental validation, researchers aim to enhance the understanding of SCWR thermal-hydraulic performance and improve the safety as well as efficiency of these advanced reactor systems. This ongoing research is critical for advancing the development of SCWRs, which hold promise for efficient and sustainable nuclear power generation in the future.

SAFARI-1 is a Materials Test Reactor (MTR) situated at Pelindaba in South Africa. SAFARI-1 is a tank-in-pool type reactor with plate-type fuel assemblies licensed to operate at 20 MW with two primary pumps. The reactor typically contains 26 fuel elements and 6 follower-type control elements. This paper will discuss the modelling and analysis approach for the SAFARI-1 reactor for normal and abnormal operation and compare results obtained against operating technical specifications (OTS) for the reactor. The operational safety of the reactor will be verified for a range of operating conditions including single failure and design bas accidents. Peak clad temperature is one of the critical parameters determining the viability of a planned cycle for the SAFARI-1 reactor. The limiting conditions of operation specify the limits for the fuel clad temperature and convection heat transfer coefficient. The maximum expected clad temperature can be calculated by performing an analysis of the neutronic and thermal-hydraulic behaviour of the proposed cycle. The transients analysed lead to an increase in fuel clad temperatures and in particular, clad temperature at the hot spot. In this analysis, onset of nucleate boiling (ONB) is used as an indicator for more in-depth analysis. The departure from nucleate boiling ratio (DNBR) is also examined for compliance.

Globally, as the development of Small Modular Reactors (SMRs) progresses, there's a corresponding surge in the development of analytical software for their design and demonstrations. This new generation of software distinguishes itself from tools used for conventional large nuclear power plant design and demonstrations by offering enhanced precision and incorporating multi-physics coupling capabilities, enabling the coupled simulation of various physical phenomena.

For instance, the United States actively utilizes the MOOSE platform in academia, industry, and research field. Similarly, Europe, China, and Japan are developing their own integrated platforms. Aligning with this global trend, South Korea launched the 'V-SMR' development project in June 2024. V-SMR, a Korean virtual reactor platform, is designed to incorporate a wide range of high-fidelity simulation software, including neutronics, thermal-hydraulics, thermal structure, and fuel performance analysis. Furthermore, it is engineered to provide supercomputing application technologies and various user-friendly features. As V-SMR is being developed to encompass various SMR analysis functionalities within Korea, it is anticipated to be utilized in the demonstration analysis of diverse reactor types in the future. This paper aims to introduce an overview of the V-SMR project and detail the key achievements made over the past year.

The System Analysis Module (SAM) is an advanced system analysis tool under development at Argonne National Laboratory, aiming to provide fast-running, modest-fidelity, whole-plant transient analyses capabilities, which are essential for fast-turnaround design scoping and engineering analyses of advanced reactor concepts. As a MOOSE-based computer code, SAM leverages modern advanced software environments and numerical methods provided by the MOOSE framework, such as its underlying meshing and finite-element library and linear and non-linear solvers. As the computer code is being widely adopted and applied in advanced nuclear reactor analyses, some numerical issues have been revealed that impact the code robustness and execution speed. Such issues could be linked to the usage of continuous Galerkin finite element method (CG-FEM) in solving thermal fluid problems. In this work, we investigate the feasibility of implementing a staggered-grid finite volume method (SG-FVM) in the MOOSE framework to support the development of advanced system analysis codes such as SAM. In this work, we demonstrated that a second-order SG-FVM is successfully implemented under the MOOSE framework as the foundation of a numerical test bed for system analysis code development. The SG-FVM-based code also exhibits superior performance in terms of execution speed based on the results of a suite of selected test problems with different problem sizes and levels of complexity. To further verify the correctness of SG-FVM-based code, it was applied to solve the Protected Loss-Of-Flow (PLOF) transient of the Advanced Burner Test Reactor (ABTR). The results show good agreement with reference results from previous studies.

The thermal-hydraulics codes DRACCAR 2024 and TRACE V5p9 simulated OECD/NEA ISP-53 tests representative of low pressure or low flowrate reflooding scenarios in Loss-of-Coolant-Accidents (LOCA). The ISP-53 tests were selected from the COAL reflooding experiments whose test section mimics PWR fuel assemblies. The objectives of the simulation exercises are to perform code-to-experiment and code-to-code benchmarks through comparisons of the simulations and the experimental results to evaluate the capability of the computer codes for modelling of reflooding thermal-hydraulics under challenging conditions such as low pressure or low flowrate.

Representative experimental results, including a few time-independent and time-dependent parameters, were chosen as figures-of-merit (FoM) to assess each code’s performance. TRACE’s simplified modelling of the rod bundle allows for faster simulations, while DRACCAR’s detailed modelling captures intricate phenomena at the expense of computational cost. The simulation results of both codes exhibited significant deviations from experimental data of the case at low pressure (3bar) and medium flowrate (50kg/m2s), with overestimated quenching speeds and underestimated peak cladding temperatures. The codes’ performance improved for simulation of the case at medium pressure (20bar) and low flowrate (17kg/m2s) although overestimation of quenching speed remained.

The low pressure or low flowrate of reflooding is a challenge for thermal-hydraulics codes to reproduce. Thus, caution should be paid when applying the codes to safety analyses of light water reactors under such conditions. The findings highlight the need for model refinements of thermal-hydraulics codes to address deficiencies in reflooding and quenching predictions, particularly for low pressure scenarios to enhance nuclear reactor safety assessments.

Micro-channel heat exchangers represent a significant innovation in heat transfer technology, offering high thermal efficiency and compact designs potentially suitable for advanced nuclear systems. Despite their potential, limited numerical analyses and experimental results are available in literature that fully characterize their performance, especially under prototypical operating conditions found in nuclear reactors. For this purpose, the Versatile Loop Facility (VLF) was designed and built to test the key components which will be part of the reactor coolant system of the Westinghouse LFR, focusing onthe Primary Heat Exchanger (PHE). The PHE is a hybrid microchannel heat exchanger manufactured using a diffusion bonding process offering a high heat transfer area-to-volume ratio, resulting in exceptional compactness, a significant advantage for the design of the primary reactor pool, as it minimizes required space allocation.

This paper presents a comparative study between RELAP5 and ANSYS-CFX for modeling microchannel heat exchangers, using the PHE of the VLF as a case study with the aim to focus on the temperature distribution, pressure drops and heat transfer coefficients and subsequently to improve the accuracy and reliability of thermal-hydraulic system codes and related modelling methodologies for design assessment and safety analyses.

Results show that the simulations of both computer codes are in a good agreement, meaning that RELAP5 provides a satisfactory overall system-level prediction.

In the reactor Emergency Core Cooling (ECC) scenario of a Pressurized Water Reactor (PWR), the external ECC coolant containing high-concentration boric acid is injected into reactor core to prevent the re-critical. The accurate estimation of the transportation of boric acid in the primary circulation is essential for the system thermal hydraulic code. The annular down-comer is the located up-stream of reactor core entrance, and the flow characteristics within it are significantly more complex than those in conventional piping. The presented work focuses on the accuracy of the boron transportation module of RELAP5 in predicting the transient boron concentration file within the annular down-comer. The experimental data that modeling the ECC scenario is introduced for the applicability evaluation. The simulation model with single-loop channel of annular pipeline component is established first. The simulation result shows the boron concentration inside the single-loop annular pipeline is almost linearly distributed, which deviates significantly from the experimental data. The improved model with four branches of annular pipeline components is proposed, where the lateral nodes of the four branches are interconnected. The improved model is capable of predicting the three-dimensional transportation of boric acid in the annular down-comer. The mean boron concentration in the four azimuthal regions of the experimental model corresponds well with the boron concentration in the corresponding branches. However, neither the conventional model nor the improved model is capable of accounting for the impact of density difference between the ECC coolant and ambient coolant on the transportation of boric acid.

Gas-liquid two-phase flow analyses are heavily involved in evaluating the safety of a nuclear power plant. System analysis codes are used to predict the system behavior of nuclear power plants. The system analysis code does not explicitly treat microscopic thermal-hydraulic behavior but uses constitutive equations incorporating its effects to achieve reliable analysis. The constitutive equations are being improved based on increasingly sophisticated measurement techniques and accumulated knowledge. The Nuclear Regulation Authority (NRA) in Japan has developed the system analysis code AMAGI as a platform to consolidate such state-of-the-art constitutive equations. NRA also continues to make extensive efforts to develop and improve critical constitutive equations, such as the drift-flux model. In this paper, the drift-flux models recently developed by the authors are implemented in AMAGI code, and the performance of the drift-flux models is evaluated by comparing experimental data and calculation results.

An analytical benchmark activity was launched within the OECD/NEA ETHARINUS project to assess the capabilities of system codes to simulate the relevant phenomena associated to the PKL Test J4.2, an Extended Loss of Alternate Power (ELAP) with the activation of the SAfety COndenser (SACO) passive system. The selected experiment allows to analyze the interactions of the primary and secondary systems with the passive system. The activity was divided into two phases: a blind phase and an open phase where participants had a period of time to improve their models. In total, 11 participants took part to the benchmark coming from a broad number of countries and applying different system codes.
The results of the blind phase showed significant discrepancies between the experimental data and the simulation outcomes, as well as among different participants' predictions. This was mainly due to the complexity of the scenario, its long duration, the strong influence of heat losses and the presence of cliff edge phenomena. During the open phase, each participant made specific adjustments to their simulation models. The challenges addressed by each participant were similar and mainly focused on improving the reproduction of heat losses and achieving a better heat balance between the systems.
The ETHARINUS benchmark provided valuable insights into the challenges of simulating complex thermal-hydraulic phenomena in nuclear reactors. The findings and recommendations from this benchmark activity contribute to enhancing the predictive capabilities of system codes and the integration of passive safety systems in nuclear reactors.

A new theory referred as Apparent Reynolds Number (ARN) has been developed to better explain the physics and mechanisms of flow laminarisation of isothermal turbulent flows caused by non-uniform body force, and then that of heated flows with strong influence of buoyancy, e.g., an upward pipe flow of air and supercritical CO₂. This concept has been extended to describe predict heat transfer deterioration in fully developed pipe air flow and now addresses mixed convection in developing air flows. In particular, the inertial terms in the momentum equations have been found to have a similar effect as the buoyancy in terms of strengthening or attenuating turbulence, leading to enhancing or deteriorating heat transfer. This understanding has prompted treating the inertia as a pseudo-body force. The ARN concept is then used to make predictions of heat transfer of developing air flow by linking turbulence mixing in complex flows such as this to that in a simple unheated shear flow based on a new equal-pressure-gradient reference framework. This has led to the development of ARN-based mixing length model. The full paper will demonstrate that this simple ARN-mixing length model can predict mixed convection heat transfer in a developing flow of air, validated against DNS data. This new physics-based modelling approach significantly simplifies the complexity of traditional turbulence models while reliably predicting complex heat transfer phenomena. It hence provides a route for modelling large energy systems with affordable computing resources. Additionally, the ARN theory enhances understanding of heat transfer behaviour in mixed convection flows.

The UK's Advanced Gas-cooled Reactor (AGR) fleet use graphite blocks for the core structure and as a moderator. The graphite undergoes dimensional change with irradiation and loses weight due to oxidation from the carbon dioxide coolant. These effects change the flow behaviour in the core and can challenge the structural integrity of the graphite bricks. Providing accurate understanding and prediction of the condition of the graphite is essential for ongoing extensions to the operational life of these reactors.

The rate of oxidation is reduced by the presence of low concentrations of other gases in the coolant. These gases need to be continuously provided to the interior of the bricks by transport of the coolant through the porous graphite material. This requires a pressure difference to be imposed across the bricks. The oxidation effects also depend on the temperature of the graphite.

To provide increased predictive insight into the flows, pressures and temperatures that influence these processes, CFD models have been built of AGR fuel channels, including all flow paths and porous flow predictions inside the bricks. The dimensions of the channel and properties of the graphite vary with the irradiation and weight loss of the bricks, which evolves and accumulates through operational life. The CFD models are able to integrate with this information coming from other parts of the analysis toolchain, and provide pressure and temperature predictions in return.

The High-Temperature Gas-Cooled Reactor (HTGR), a proposed Generation IV nuclear reactor, is gaining increasing attention for its inherent safety, high thermal efficiency, and ability to produce high-temperature process heat. The successful deployment of the HTGR technology depends on an in-depth understanding of reactor physics, particularly coolant flow, heat transfer within fuel assemblies, and their impact on reactor structural integrity. While Computational Fluid Dynamics (CFD) can provide detailed 3-D predictions of the thermal-hydraulic behaviour in the reactor core, the large computational resources required make it impractical for real-world nuclear engineering applications.

This work presents a coarse-grid CFD approach, initially developed for light water reactors, which has now been extended to prismatic HTGR fuel assemblies. This method, known as Subchannel CFD (SubChCFD), combines the strengths of traditional subchannel codes and modern CFD. It offers CFD-like 3-D predictions for a large range of scenarios, and meanwhile, the results produced are consistent with well-calibrated empirical correlations. By using a coarse mesh, SubChCFD reduces the computing costs by up to 1 to 3 orders of magnitude compared to conventional Reynolds Averaged Navier Stokes (RANS) CFD, depending on the complexity of the problem. This potentially makes the full reactor core simulations more feasible and cost-effective. To demonstrate the versatility of SubChCFD, the General Atomics modular HTGR fuel assembly is investigated. The results show that SubChCFD simulations of a full-length prismatic HTGR fuel assembly closely align with conventional RANS simulations for the same problem, but the computational cost is significantly lower than the latter.

The Single Primary Heat Extraction and Rejection Emulator (SPHERE) facility at Idaho National Laboratory was recently utilized to generate data for the startup and steady operation of a high-performance, sodium heat pipe over the course of 1000 hours, as a test of detrimental, long-term effects of heat pipe operation. The setup consists of a single, sodium heat pipe enclosed in a stainless-steel vacuum chamber, heated radiatively via a cylindrical ceramic fiber heater configuration and cooled via a water-cooled calorimeter. Measurements include temperatures at several axial locations along the outer surface of the heat pipe, the power provided to the heaters, and the heat removal rate of the calorimeter. In this work, this data is utilized to validate heat pipe models in the heat pipe application Sockeye, which is based upon the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. Sockeye provides various heat pipe models at an engineering scale appropriate for the multiphysics simulation of microreactors, which may feature several hundred heat pipes. This work details models of this experiment in SPHERE using various heat pipe models with Sockeye, including heat conduction-based models and compressible flow models of the heat pipe interior. These models are compared to the experimental data to assess the accuracy of several aspects of heat pipe modeling, including frozen startup, the effect of non-condensable gases, and the coupling of the heat pipe to its environment.

Due to their compact structure and strong mobility, solid-fuel heat pipe reactors have gradually become a research focus for small reactors. Current research mainly concentrates on nuclear-thermal-structural multi-physics coupling, considering thermal expansion. However, during the long-term operation of heat pipe reactors, the reactivity feedback caused by fuel irradiation-induced swelling must be considered. Therefore, based on RMC and OpenFOAM, this paper develops an analysis process for nuclear-thermal-structural multi-physics coupling in solid-state heat pipe reactors, taking irradiation effects into account, and conducts a study on the KRUSTY heat pipe reactor. The results show that for KRUSTY, due to the low burnup, the negative feedback from irradiation effects is not as significant as that caused by thermal expansion. However, the overall calculation indicates that irradiation effects must be considered for reactors with high burnup.

System-level modeling of the 200 MW General Pebble Bed Reactor (GPBR200) is performed with SAM. Using SAM’s component-based system, a core channel approach is developed and used to model the core of the GPBR200. For comparison, a SAM porous-media multi-D model is also developed for the same reactor. Good comparisons are obtained for the two model during steady-state normal operating condition. Furthermore, transient simulations are performed for the de-pressurized and pressurized loss-of-forced cooling accidents (DLOFC and PLOFC). The core channel model compares well with the porous media model during DLOFC but overpredicts the overall temperature of the reactor during PLOFC. The good comparison during DLOFC indicates that the core channel model is able to capture radial conduction well. On the other hand, the overprediction of temperature by the core channel model during PLOFC suggests that the model underestimates the effects of in- core natural circulation during the transient.

Sodium-cooled fast-neutron reactors (SFR) are currently considered to be the most mature type of reactor able to optimize uranium ore usage and reduce nuclear waste produced from Generation II and III reactors. CEA led studies up to 2019 on the features of a 600MWe reactor within the frame of the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project. The chosen pool-type design offers the advantage of containing the primary sodium within a single vessel, ensuring safer operations by transferring heat to the secondary sodium circuit via Intermediate Heat Exchangers (IHX). This design eliminates the risk of water/primary sodium interaction. A tertiary loop then generates steam for power conversion. Given the safety implications of the design, careful study of the vessel's geometry is essential, particularly the IHX, which plays a critical role in heat exchange. To investigate the flow dynamics within the vessel, a scaled-down model of the ASTRID reactor was constructed. Using a similarity approach water was used as a simulant fluid due to the complexity and cost of sodium-based experiments. This model allows for adjustments in IHX geometry to conduct parametric studies on flow behavior. Particle Image Velocimetry (PIV) was employed to measure velocity near the IHX inlet across different configurations. The results align with previous studies, indicating that, whatever the configuration, flow is concentrated in the lower section of the IHX, offering valuable insights for future design improvements.

This paper presents a safety analysis performed using the AMESIM (Advanced Modeling Environment for Simulation of Engineering Systems) code for the PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor), proposing an appropriate methodology for global export market. The safety analysis for the PGSFR has been carried out with the MARS-LMR code. This research aims to develop a transient safety analysis platform for SMR (Small Modular Reactor)-powered ships. The AMESIM code offers advanced numerical methods capable of solving complex multi-physics problems, making it suitable for modeling not only thermal-fluid systems but also mechanical and electrical systems, thus fitting the modeling of ship propulsion systems. However, there are challenges in modeling nuclear systems with AMESIM. Therefore, this study defined coolant properties and modeled reactor systems of the PGSFR in AMESIM to evaluate the applicability of nuclear systems in the AMESIM SW environment. In the AMESIM code, the PGSFR consists of the Core, PHTS (Primary Heat Transport System), IHTS (Intermediate Heat Transport System), and SG (Steam Generator). In the Core, Reactivity Feedback and Point Kinetics are calculated to determine the Neutron flux. It was found that the results from AMESIM code have a good agreement with design values of the PGSFR. Furthermore, preliminary safety analysis for representative DBEs (Design Bases Events) in a PGSFR has been implemented with AMESIM code.

To enhance the safety of sodium-cooled fast reactors, the decay heat removal system under natural circulation with a dipped-type direct heat exchanger (D-DHX) installed in a hot pool of a reactor vessel (RV) has been investigated. During the D-DHX operation, the thermal-hydraulics of RV is complicated because the cold sodium from the D-DHX flows into the core and the radial heat transfer among assemblies occurs. To evaluate the RV thermal-hydraulics and core cooling performance given from these phenomena in the design study, we have been constructing the RV model using a computational fluid dynamics code (RV-CFD) with the subchannel CFD (SC) model for assemblies as a practical model which can achieve a lower computational cost while maintaining prediction accuracy (RV-CFD). However, the applicability investigation of RV-CFD was limited to several numerical analyses of steady-state. In this study, to evaluate accurately the transient response of sodium temperature using the RV-CFD, we develop the non-equilibrium thermal (NET) model in the SC model which can consider both the heat capacity and thermal resistance in simulated fuel pins. The transient analysis simulating the power reduction due to reactor scram from the steady-state operation in a sodium experimental apparatus named PLANDTL-1 is conducted. The result shows the thermal-hydraulic behavior in the RV during the transient is predicted, and the core temperature in the transient is reproduced. Thus, the RV-CFD using the NET model in the SC model can evaluate the transient temperature response.

Ensuring the safety of liquid metal-cooled reactors necessitates accurate modeling of the transition from steady-state operation to long-term passive cooling under various initiating events. A significant challenge exists in a protected loss of flow event, where thermal stratification developing in the reactor pools can impact or delay the transition to long-term cooling through natural circulation. This can induce unexpected thermal gradients which can lead to oscillating temperature fields resulting in off-normal thermal-hydraulic behavior throughout the system. This paper describes ongoing activities at Argonne National Laboratory to validate system-level software using the Thermal Hydraulic Experimental Test Article (THETA) of the Mechanisms Engineering Test Loop (METL) to enhance the system-level tools used to assess safety margins. The experimental campaign using THETA, designed to operate at scaled-down prototypical pool-type sodium-cooled fast reactors (SFRs) conditions, has been evaluated to expand the validation basis for modeling thermal stratification during and after the transition to natural circulation. A system-level computational model using SAS4A/SASSYS-1 has been developed to represent the full THETA facility, including the modeling for the facility electric heater, primary and secondary pumps, an intermediate heat exchanger, an air-cooled heat exchanger, hot and cold pools, and connected piping across both the primary and secondary systems. The THETA SAS4A/SASSYS-1 model uses a stratified volume model for both the hot and cold pools with heat transfer interactions across major components. The preliminary assessment results are discussed with key findings and potential directions for improvements of the model.

With the growing global interest in small modular reactors (SMRs), one of the key goals of the new European ESFR-SIMPLE project is to develop a compact sodium-cooled fast reactor (SFR). This system aims to address crucial SMR features, such as the transportability of main components and grid flexibility, while also leveraging the extensive experience in sodium coolant technology. Additionally, reducing core power could enhance safety by improving inherent reactivity characteristics and enabling more efficient removal of residual power, potentially paving the way for constructing a prototype SMR-SFR in Europe.

This study presents the initial results of primary system modeling for the SMR-SFR with a thermal power of 360 MW. The ATHLET system code was used to simulate the sodium coolant flow in the primary system, exploring various modeling options. Notably, the application of models for pseudo-3D flow in the large hot plenum of the primary vessel was of particular interest. The paper discusses the simulation results for selected transients and the findings from comparing the conventional zero-dimensional plenum approach with alternative pseudo-3D flow modeling options for the hot pool.