Conference Agenda

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.

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.

An Annular Linear Induction Pump (ALIP) was designed for driving Lead Bismuth Eutectic (LBE). The basic parameters of the ALIP were calculated by the multi-physics coupling software COMSOL. The ALIP have 4 pole pairs, a frequency of 50 Hz, an input line current ranging from 0 to 80 A, and a corresponding output head ranging from 0 to 500 kPa, with a flow rate of 0 to 10m³/h. Experiments were conducted within the current range from 28 to 52A, the results showed that the experimental values matched well with the calculated values. Experiments on the output head of the ALIP was conducted with LBE at temperatures of 250, 300, and 350℃. The results showed that the output head of the ALIP varied little under the same electromagnetic parameters. This is due to the small change in the resistivity of the LBE with temperature, which is significantly different from sodium. The head curve of the ALIP was tested at a LBE temperature of 300℃ by adjusting the input electromagnetic parameters. The results indicate that the output head and LBE flow rate of the ALIP increase with the increase of input current, voltage, and power. However, under the same input electromagnetic parameters, the output head of the ALIP decreases as the flow rate increases.

The development of pool-type reactor MYRRHA, utilizing heavy liquid metal coolant necessitates primary pumps with extended massive shafts supported below the coolant free-surface level. Hydrostatic bearings are the most suitable choice for these specific conditions. However, conventional calculation methods for hydrostatic bearings are inadequate for the unique operational parameters presented by this application. This study focuses on the computational fluid dynamics (CFD) analysis and optimization of hydrostatic bearing designs for primary pumps in MYRRHA. Three bearing design candidates with different numbers of pockets were initially evaluated using CFD simulations on a scaled-down test model of the primary pump. The most promising design underwent iterative refinement to meet specific performance requirements, including pressure drop, load capacity, pressure ratio, and frictional torque. A comprehensive parametric analysis was conducted on the optimized design to characterize its performance across various operational scenarios, including the study of the influence of rotational speed and eccentricity. The CFD model developed for this analysis incorporated mesh optimization and turbulence modelling, simulating heavy liquid metal flows in the restrictors, the pockets, and the narrow gap of the hydrostatic bearing. The outcome of this research is a hydrostatic bearing design that satisfies all specified requirements for use in the scaled-down test model of the primary pump of MYRRHA. The CFD modelling approach provides a robust and reliable framework for future design and optimization efforts in this field, contributing to the advancement of primary pump hydrostatic bearing technology in heavy liquid metal-cooled reactors.

Amongst Generation IV reactor designs, liquid metal-cooled reactors boast high power density owing to the high thermal conductivity of metals. The thermo-hydraulic phenomena that occur in the reactor pool in different scenarios (steady operation, accidents, non-critical transients, etc.) are a topic of great interest in the research community. One such reactor concept is MYRRHA, a flexible fast-spectrum research reactor cooled by lead-bismuth eutectic (LBE) under design at SCK CEN. To support the design of MYRRHA and provide data that gives insight into such phenomena and for numerical code validation, a 1/6th scale model called European SCAled Pool Experiment (E-SCAPE) was developed.

As part of the European project PASCAL, NRG aims to perform multi-scale simulations of E-SCAPE subjected to asymmetric accident scenarios of Heat Exchanger and Single Pump Failure coupling the in-house System Thermal Hydraulic (STH) code SPECTRA to the commercial Computational Fluid Dynamics (CFD) code STAR-CCM+ via the in-house coupling tool myMUSCLE: MultiphYsics MUltiscale Simulation CoupLing Environment.

In previous articles, the standalone as well as coupled models of E-SCAPE have been validated against a steady state isothermal scenario. In this article, the next step is taken by performing coupled calculations of the steady active operation at a mass flow rate of 93.2 kg/s and 80% power, i.e 73kW, that serves as the pre-accident state to the Heat Exchanger Failure scenario, and comparing to the standalone STH and experimental results. The calculations reveal stable solutions that are well in agreement with the standalone STH as well as the experimental results.

Lead(Pb) and lead bismuth(Pbbi) reactors are potential types of fourth generation reactors. Lead reactors have higher thermal efficiency and natural circulation capabilities. At present, there are almost no publicly reported experimental data on the heat transfer characteristics of liquid lead, especially the data under natural circulation mode. In this study, a pool-type natural circulation experimental platform was first designed, which includes a simulated core(simulated with 37 heating rods with a length of 1200mm, and P/D of 1.3), a hot pool, a cold pool, upper and lower channels, and four symmetrical lead-oil heat exchangers. During the experiment, liquid lead undergoes endothermic expansion in the simulated core and flows into the top lead-oil heat exchangers through a hot pool. After heat exchange, the liquid lead flows downwards along the cold pool into the bottom of the simulated core, completing natural circulation. The T/H characteristics of liquid lead at simulated core, hot pool, cold pool, etc. were analyzed and studied. At the same time, experimental modeling based on system code was also carried out, and the experiments were compared and verified with the code. The research results can provide support for the design of liquid lead pool-type natural circulation reactors.

System codes can efficiently handle system-level problems and obtain transient characteristics of whole system. However, they lack the ability to analyze the local flow and heat transfer characteristics of components. CFD codes have the ability to perform highly precise analysis of local components, but cannot analyze the whole system. Therefore, it is an important direction of current research that achieving the coupling calculation of system and CFD codes. To obtain the flow and heat transfer characteristics of the core and transient response of the lead-bismuth fast reactor, a coupling code between system code and CFD was developed. Through a data transferring platform based and explicit coupling method, the simulation of the primary loop of lead-bismuth fast reactor core was achieved. To verify the coupling code, system code and coupling code under the same operating conditions were performed. The flow rate and coolant temperature in the primary loop of the lead-bismuth fast reactor were compared. It was found that the coupling simulation results were consistent with the results of system code, indicating that the coupling code can accurately predict the flow and heat transfer characteristics and system response of the lead-bismuth fast reactor core, and verify the feasibility and rationality of the coupling method. This study provides a coupling method for the thermal hydraulic analysis of lead-bismuth fast reactors.

The artery alkali-metal heat pipes in reactors are essential for energy transfer, with dynamic thermal performance, such as two-phase circulation startup and capillary heat transfer limits, posing challenges to overall reactor performance. Investigating the capillary dynamics behind the transient thermal load is crucial for understanding the operational characteristics of artery heat pipes. This work aims to investigate the complex heat and mass transfer phenomenon of the capillary limit, which is characterized by dynamic non-equilibrium and multi-scale and multi-physical coupling, by conducting this research from the three dimensions of micro-mesoscopic mechanism, macro heat transfer characteristics, and reactor system operating patterns. By developing a theoretical framework for dynamic capillary heat transfer, a dynamic thermal analysis model for the artery heat pipes has been established and validated through experiments. The average error of the capillary dynamics model compared to experiments is 3%, while the dynamic heat transfer model shows less than 10% error against CFD simulations and under 20℃ error compared to steady-state and transient experimental results, confirming the model's accuracy. Additionally, the study investigates the correlation between capillary dynamics and dynamic heat transfer phenomena, identifying three startup phases of free molecular flow, continuous flow expansion, and continuous flow. It categorizes capillary limits into gas-phase and liquid resistance-dominated types based on two-phase countercurrent circulation. By combining the weak feedback characteristics of fast reactors with the dynamic heat transfer of artery heat pipes, the study proposes operational strategies for typical heat pipe reactors, examining system behavior during startup and under transient thermal loads.

The focus is on a 10 MW thermal Advanced High Temperature gas-cooled Micro Reactor (AMR) currently being designed. The reactor will employ prismatic graphite blocks for structural and moderator material. There will be 420 fuel assemblies in the core using low enriched TRISO fuel contained in borings within the fuel graphite blocks that allow annuli for cooling. The thermal-hydraulic behaviour of the reactor under steady-state conditions and during a Depressurized Loss of Forced Cooling (DLOFC) event has been simulated employing an axi-symmetric systems network model using Flownex SE. Under steady-state conditions the helium coolant enters the reactor at 320 C and exits at 750 C. It is found that the bottom of the core is 403 C hotter than the top of the core and in the radial temperature gradient is distorted due amongst others to an average drop in temperature of 220 C between the last fuel ring and the outer reflector (OR). The OR transfers 618 kW to the coolant flowing up the risers placed in the OR, preheating the coolant 346.8 C. The reactor cavity cooling system (RCCS) rejects 86.1 kW. During the first 5 seconds of the DLOFC the mass flow rate through the initially increases due to the blowdown effect, and the heat transfer to the also fluid increases initially. Subsequently the heat rejected by the RCCS reach a maximum of 109 kW. It found that the heat released by the solids can constitute up to 45.5% of the heat rejected by the RCCS.

The eVinci™ Microreactor which is under development by Westinghouse Electric Company could bring a cost-competitive and reliable nuclear energy source to the world. The small size of the eVinci microreactor allows for transportability and rapid, on-site deployment.

Instead of a fluid-based primary coolant system normally seen in nuclear power plants, eVinci Microreactor adopts heat pipes to transfer heat from the reactor to the Primary Heat Exchanger (PHX). The heat pipe design enables passive core heat removal which eliminates numerous components needed in active coolant systems and makes the eVinci microreactor a pseudo “solid-state” reactor with minimal moving parts.

The eVinci Nuclear Test Reactor (NTR) is a nuclear test facility dedicated for eVinci microreactor’s development. The NTR will provide critical engineering information for analysis code validation to support commercial licensing. A CFD model has been developed to support NTR PHX design optimization. A two-step method was employed for the NTR PHX CFD modelling. Step 1: A series of cases for single heat pipe finned-sleeve tube were simulated with the finned channel simplified as a porous media. The expressions for resistance and heat transfer coefficient were derived for porous media. The results were benchmarked to the test data. Step 2: Applied the derived expressions for porous media parameters from Step 1 to a full PHX CFD model. The results from Step 2 were used to help PHX design optimization. In this paper the two steps of the NTR PHX CFD model development were presented.

In this paper a multiphysics integrated full-core 3D model and the analysis results of the Westinghouse Nuclear Test Reactor (NTR) are presented, coupling the neutronic and thermal analysis in the reactor core and the heat transfer in the primary heat exchangers. The NTR reactor is an advanced 2~3 MWt transportable heat pipe cooled microreactor currently developed by Westinghouse. It is an epithermal reactor with prismatic solid core using TRISO particle fuel embedded in cylindrical fuel compacts.

The software tools used are finite element (FE) solvers developed in the framework of the Multiphysics Object Oriented Simulation Environment (MOOSE) under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program sponsored by the US Department of Energy (DoE). MOOSE-based multiphysics modules of neutronics and thermal-hydraulics are coupled in solving the 3-D fission power and temperature distributions in the full core NTR reactor model. Furthermore, the reactor core model is coupled with 1-D flow models of the cooling air channels over the condenser sections of heat pipe, simulating the heat transfer in the Primary Heat Exchanger (PHX). Non-uniform flow and inlet temperature among air flow channels are informed by detailed computational fluid dynamics (CFD) calculation of the PHX. Using this integrated model, several Design Basis Accidents (DBAs) identified for the NTR design are simulated, including the accidents initiated from inadvertent control drum rotation (reactivity insertion), total loss of PHX, and single heat pipe failure.

The Single Primary Heat Extraction and Removal Emulator (SPHERE) facility at Idaho National Laboratory (INL) has been instrumental in advancing the development and validation of heat pipe technologies for microreactor applications. As a part of these efforts, long-duration heat pipe tests are required to assess long-term reliability concerns related to wick degradation, corrosion, manufacturing methods, and compatibility of materials. This paper presents the findings of a long-duration test conducted at the SPHERE facility, focusing on the performance and reliability of a high-performance, defined as over 2kW, heat pipe under sustained operational conditions. The tests emulated the anticipated common thermal characteristics of microreactor concepts. The results show the robustness of heat pipes, with a significant amount of data collected on the ratio of heat losses to heat transported and degradation rates over an extended period. Key data and performance metrics, including time series of temperatures, axial temperature profiles, thermal response times, and heat transfer capabilities, the thermal output over thermal input, were reported and discussed. These findings provide critical insights into the design and optimization of heat pipes, underscoring their potential to enhance the safety and efficiency of next-generation reactor concepts.

Heat pipes are two-phase heat transfer devices that enable passive removal of heat from the reactor core to the power conversion system in heat pipe-cooled microreactor designs. Experimental investigations of heat pipe transients are needed for technology demonstration, verification and validation of numerical codes, and the establishment of regulatory requirements. The Single Primary Heat Extraction and Removal Emulator (SPHERE) facility at Idaho National Laboratory (INL) serves as a platform for evaluating the dynamic response of high-temperature heat pipes under a variety of operating conditions. The present work details the experimental investigation of a high-performance, defined as over 2kW sodium heat pipe subjected to rapid input power fluctuations induced by sudden changes in the evaporator temperature setpoint. In addition, the heat pipe was subjected to an asymmetrical heat load where a subset of heaters operated at 30% and 70% below their nominal power. These experimental conditions were chosen to simulate thermal and operational stresses expected to be encountered in microreactors to provide data on heat pipe behavior during such important transient events. Key data and performance metrics, including time series of temperatures and strains, axial temperature profiles, thermal response times, and heat transfer capabilities, the thermal output over thermal input, were reported and discussed. The results highlight the resilience of heat pipes, revealing their potential to maintain thermal stability and efficiency under varying power loads. Lastly, the paper concludes with a discussion on the significance of the results and their implications for future research.

This study enhances the understanding of thermal effects on energy distribution in the near-wake region of flow over a heated sphere by analyzing time-resolved particle image velocimetry (TR-PIV) experimental data at elevated pressures (3 MPa). The experiment spans a wide range of Reynolds numbers (19,000–29,000) and Richardson numbers (0.5–2.0), conditions characteristic of opposed flow mixed convection. Key parameters, including mean and fluctuating velocities, were calculated from the acquired velocity vector fields, with uncertainties quantified. The unique contribution of this work lies in examining the lateral and streamwise expansion of the recirculation region as heating increases, and in comparing these results with isothermal conditions. Additionally, this study isolates the effects of natural convection by comparing time-resolved turbulent kinetic energy (TKE) at the streamwise center of the recirculating region for both heated and unheated cases. Spectral analysis was conducted on the Reynolds-decomposed streamwise and spanwise velocity components using Power Spectral Density (PSD), providing insights into the turbulence characteristics within and outside the wake region. These findings are particularly relevant to the design and safety of Pebble Bed Gas-Cooled Reactors (PB-GCRs) due to the similarity in geometry and operating conditions. This work contributes to advancing the understanding of mixed convection in nuclear reactor cooling systems, offering insights into thermal-hydraulic performance under elevated pressures and varied thermal conditions.

Pronghorn is a thermal-hydraulics computational tool developed using the Idaho National Laboratory's Multiphysics Object-Oriented Simulation Environment (MOOSE). It is designed to support Computational Fluid Dynamics (CFD) modeling, ranging from subchannel and porous media analysis to Reynolds Averaged Navier-Stokes (RANS) turbulence modeling. As an integral part of the MOOSE-based suite of tools, Pronghorn seamlessly couples with other MOOSE-based applications to simulate a variety of physical phenomena. This article highlights recent significant enhancements to Pronghorn's CFD modeling capabilities and demonstrates their application to advanced nuclear reactor designs.

The recent improvements in Pronghorn primarily focus on modifications to its turbulence modeling capabilities, near-wall corrections and numerical schemes. In terms of turbulence modeling, the two-equation k-ϵ and k-ω-SST models have been implemented and validated with both equilibrium and non-equilibrium wall treatments. Regarding numerical schemes, a second-order hybrid method for gradient computation has been developed, resolving issues related to solution artifacts on skewed computational meshes commonly found in the complex advanced reactor designs. Additionally, corrections for wall roughness, and curvature, and wall-channeling in pebble beds have been introduced in the near-wall modeling.

These developments enable more accurate simulations of advanced nuclear reactors. Two case studies are presented in this work: a pool-type Molten Chloride Reactor and a salt-cooled Pebble-Bed High Temperature Reactor. In both cases, the previous models in Pronghorn are compared with the new implementations, demonstrating the improved accuracy achieved with the updated models.

OECD-NEA has launched thermal hydraulic code validation benchmark for high temperature gas-cooled reactors using the High Temperature Test Facility (HTTF) data. KAERI has joined as a participant to compare the calculated results by GAMMA+ code with the experimental data and code-to-code. Based on full power operation assumption, the steady state temperature profiles by the different codes were compared. During the initial comparison process, it showed that the calculated results of each participant were slightly different. It was thought that it could be from different nodalization and modelling approaches. The feature of the real HTTF test has prismatic blocks with fuel compact holes and coolant holes. But, the computational nodes by the each system code were simplified due to limitation of system codes as equivalent cylindrical domain. Several modelling methods to the each radial node has attempted to get more close data. Steady, PCC and DCC events were analyzed with the best acceptable method in this benchmark.

The impact of optical errors from physical sources was studied to assess their influence on PIV (Particle Image Velocimetry) measurements and 3D photogrammetry reconstructions of pebble beds. This investigation provides guidelines for optimizing these physical parameters to ensure successful optical measurements in "large" pebble beds. In this context, a large pebble bed refers to one containing more than 1,000 pebbles, with a length of at least 10 pebble diameters in each direction. Previously published studies that use similar techniques have pebble beds up to 1000 pebbles in size, but with a depth of around 7 pebbles in the narrowest direction.

Alongside the error analysis results, preliminary PIV measurements are presented, including details on calibration methods and other aspects crucial for generating CFD-grade experimental data. This data is essential for validating CFD tools like NEK-RS, which are progressively improving toward fully resolving the flow dynamics within full-scale PBR (Pebble Bed Reactor) pebble beds. Finally, results from creating a 3D reconstruction of the experimental pebble bed will be discussed. This reconstruction is both as challenging as the optical flow measurements and equally important for generating high-resolution data relevant to simulation validation.

In the reactor core of HTR-PM, due to the structural materials such as graphite blocks and carbon bricks arranged in bulk, the coolant flow paths are complicated. A part of the coolant flows through narrow gaps between the structural materials without cooling the pebble bed, which affects the temperature distribution in the reactor core. Therefore, the accurate simulation of bypass flow is a key issue related to reactor safety. The vertical gaps between side reflectors are the main bypass flow paths.In this paper, the thermal hydraulic system code was employed to simulate flows in the pebble bed and vertical gaps, analyze the flow path of coolant under different bypass flows ratio, and explored the influence on the temperature distribution of the pebble bed and the side reflectors.The numerical results are proven to be in good agreement with experimental data and results by CFD. The model reasonably simulates the bypass flow of core coolant and the temperature distribution in the core. The results also shows that there are some flow paths different from the previous researches and there is the flow direction turning point of bypass flow in vertical gaps.This method solves the problem of high computational cost when using the CFD method to study bypass flows. It is able to calculate accurately while greatly reducing computing costs, which lays a good foundation for further safety analysis under accidents.

This study conducts multiphysics modeling of graphite prismatic blocks used in High Temperature Gas- cooled Reactors (HTGRs) by analyzing the mechanical and thermal property changes induced by neutron irradiation. Graphite conducts a critical role as a moderator, reflector, and structural material in HTGRs, and these properties are significantly dependent on the level of radiation exposure in high temperature and neutron irradiation environments. The radiation-induced creep and dimensional changes have a substantial impact on the structural stability of graphite components, making their assessment essential.

Through 3D structural simulations, insights into the mechanisms of creep stress and dimensional changes occurring at elevated temperatures are provided, enhancing the understanding of how these changes affect the structural stability of graphite. The stress analysis results including this creep phenomenon are expected to be fundamental for evaluating the failure probability of the graphite prismatic block designs.

Accurate prediction and assessment of core bypass flow are vital, as they affect the heat transfer and cooling efficiency of the reactor. To address this, coupled CFD and mechanical studies considering neutron irradiation and thermal expansion have been conducted. The volume expansion with neutron irradiation dose decreases the width of the bypass gap, which increases the pressure drop but increases the heat transfer efficiency by the coolant hole.

This research is expected to contribute to the reliability evaluation of graphite components in HTGRs and provide insights for future reactor core designs and operation, enhancing the stability and efficiency of helium cooling under radiation.

A generic Small Modular Reactor (SMR) simulation model was developed in two system codes: Apros and TRACE. NuScale design data and other public SMR design data was used as a reference point for the development of the model. The objective of the work was to study the modelling choices and simulation capabilities of the selected codes with respect to the SMR design features (e.g. natural circulation systems, helical coil steam generator, compact vacuum containment, reactor pool and integrated pressurizer). In particular, the goal was to assess the suitability of Apros and TRACE simulation codes for the simulation of SMR applications. This was done by calculating one steady-state and three transient simulations (inadvertent operation of emergency core cooling system and two variations of turbine trip) with the developed simulation models. The results were compared with the reference simulation results presented in NuScale final safety analysis report (FSAR) to assess the capability of the codes and suitability of the modelling choices. One of the turbine trip cases was also compared with two previously published reference results by two other codes to complement the comparison and provide insight into the analysis of the results. Good match with the overall trend of the reference results was achieved with both Apros and TRACE simulation models which confirms the capability of the codes to model this type of SMR configuration and simulate both steady-state and typical transients.

This study experimentally investigates the frictional pressure loss and flow regime behavior of a 9×9 mock Helical Cruciform Fuel (HCF) rod bundle, a novel design proposed as a potential replacement for conventional cylindrical rods in Light Water Reactors (LWRs). The unique cruciform cross-section, featuring four twisted petals, eliminates the need for conventional spacer grids, offering higher fuel packing fraction and enhanced coolant mixing. To assess these advantages, a high-precision differential pressure measurement system was employed over a Reynolds number range of 200 to 22,000, covering laminar, transition, and turbulent flow regimes. The experimentally determined friction factors showed statistically similar trends between the “one pitch” and “bundle-averaged” axial segments, confirming fully developed flow in both regions. Empirical correlations for friction factor and differential pressure per unit length were then developed for each flow regime and validated by comparison to previous HCF and wire-wrapped fuel bundle studies. Results identified flow regime boundaries at approximately Re ≈ 1,000 for laminar-to-transition and Re ≈ 8,274 for transition-to-turbulent, highlighting distinctly different hydraulic behavior in the three regimes. The findings significantly broaden the limited experimental database on HCF rod bundles, providing new insights into regime-dependent pressure drop characteristics. By refining existing correlations and offering high-fidelity benchmark data, this work advances the development of more efficient and accurate reactor core designs that leverage HCF technology for enhanced thermal performance.

NuScale is a light water cooled small modular reactor with an integral reactor pressure vessel design that relies on natural circulation to provide the primary mass flow. This work focuses on the simulation of LOCA sequences caused by a break in the CVCS discharge line inside the steel containment. For this purpose, a model of NuScale was developed using the TRACE system code, which includes modeling of the primary and secondary systems, the steel containment, the reactor pool, and the safety systems. In this study, the base case corresponds to a LOCA in which the ECCS fails without opening either of the reactor recirculation valves. This scenario is selected based on the PRA results included in the NuScale DCA. A sensitivity analysis is then performed to determine the time available to manually actuate CVCS injection. Further simulations were also performed with the recovery of one out of two RRV openings. The results allow comparison of the time available for each LOCA management strategy.

In recent years, development of novel passive safety systems for new reactor designs has significantly increased. These systems are recognized for their ability to operate reliably for extended period of time and without the need for operator action or active components requiring electricity. The Innovative Small Modular Reactor (i-SMR), an integral-type SMR that is currently being developed in South Korea, incorporates two such passive safety systems. The Passive Auxiliary Feedwater System (PAFS) is intended for long-term core cooling and decay heat removal by condensation of steam removed from the steam generator. The Passive Containment Cooling System (PCCS) is designed to depressurize the containment vessel during a Loss of Coolant Accident (LOCA), replacing the conventional Containment Spray System (CSS). The performance of both PAFS and PCCS is governed by a heat transfer driven by a two-phase natural circulation flow, presenting several design challenges. Traditional deterministic safety assessment using system codes lack the precision needed to capture the detailed dynamics of phenomena occurring within passive safety systems, such as rapid steam condensation and associated multi-dimensional flow. Accurate prediction of the PAFS and PCCS performance under accident conditions necessitates a thorough understanding of these behaviors. This study therefore leverages the MULTID component for reliable simulation of the dynamic three-dimensional phenomena associated with operation of the passive safety systems, along with the overall plant response, evaluated using the MARS-KS. The main focus of this study is the EDV-LOCA scenario, where both PAFS and PCCS play a crucial role for effective accident mitigation.

Westinghouse AP300^TM SMR is the latest Westinghouse small modular reactor based on the proven AP1000® pressurized water reactor technologies to accelerate its development and deployment. The passive safety system of AP300 is the same but scaled down from the industry leading AP1000 passive safety system, which has been extensively analyzed and tested. The testing basis of AP300 is expected to be well covered by the extensive AP600/AP1000 testing programs, which consists of many separate effects test facilities and integral effects facilities for both passive core cooling system and passive containment cooling system, such as APEX600/1000, ROSA-AP, SPES, Madison CMT test, VAPORE, PRHR HX test, LST, PCS water distribution test, condensation test, etc. In addition, the program also includes the previous large break LOCA experiments that are essential for the licensing of AP600/AP1000 such as UPTF and CCTF experiments. These experiments will be reviewed and the applicability of the facilities and the experiments to the AP300 SMR will be discussed.

The Generation IV (Gen-IV) International Forum is a framework for international co-operation and collaboration in research and development for the next generation nuclear energy systems. Within the sodium-cooled Fast Reactor (SFR) system arrangement, there are four projects; System Integration Assessment (SIA), Advanced Fuel (AF), Component Design & BOP (CD&BOP), and Safety & Operation (SO). The SFR SO project addresses the areas of safety technology and reactor operation technology developments. It aims for (1) analyses and experiments that support establishment of the safety approaches and validate the performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFRs. The tasks in the SO area are categorized by the following three work packages (WP). WP-SO-1 "Methods, Models and Codes" is for the development of tools used to evaluate the safety. WP-SO-2 "Experimental Programs and Operational Experience" is for the operation, maintenance and testing experiences in experimenta facilities and SFRs. WP-SO-3 "Studies of Innovative Design and Safety Systems" is for safety technologies of Gen-IV reactors such as active and passive safety systems and other specific design features. This paper includes recent activities of member countries and organizations within the SFR SO project.

Liquid metal cooled fast reactors (LMFR) use liquid metal as the primary coolant of the reactor core. First demonstrated in the 1950s, they were never fully deployed compared with the light water reactor technologies. However, the early 2000s saw a resurgence of interest, particularly in Sodium Fast Reactors (SFR) and Lead Fast Reactors (LFR) as Generation IV designs, due to their potential to significantly reduce the amount and toxicity of nuclear waste in a closed fuel cycle. This investigation is part of Tractebel’s effort to evaluate new tools for both SFR and LFR modeling. CTF, a subchannel thermal-hydraulic code for Light Water Reactor applications, has been used at Tractebel since 2015. It incorporates state-of-the-art models, correlations, and methods for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) modeling. Recently, new features have been developed in CTF to model SFR and LFR reactor cores. Given Tractebel’s expertise with the code, CTF is a promising candidate for developing LMFR modeling capabilities. This study focuses on validating CTF models for SFRs using data from two test facilities: TAMU 61-pin isothermal tests (Texas A&M University and SEFOR (Consortium of Southwest Atomic Energy Associates, Karlsrühe Laboratory, Euratom, General Electric). This data is obtained through participation in OECD/NEA benchmarks LMFR T/H and SFR-UAM. Key models of interest include friction factor correlations, turbulent mixing, and Nusselt correlations for heat transfer in liquid metals. This paper presents the preliminary outcomes of these investigations.

The use of Separate Effect Tests (SET) and Integral Effect Tests (IET) is a common practice in support of Sodium Fast Reactors (SFR) designs. These tests are built to analyse physical phenomena and their measured data can serve as validation database for simulation codes. In the framework of the Franco-Japanese collaboration on Research and Development for SFR thermal hydraulics, transient tests were performed in the IET named PLANDTL2 test facility in Japan. This IET’s instrumented test section is composed of an electrically heated core and a hot pool with a Dipped Heat Exchanger (DHX). The Intermediate Heat Exchanger (IHX) and the Electro-Magnetic Pump (EMP) are located in a deported primary loop. Studied transients consist in transition from forced convection to natural convection, in the pool and in the primary circuit, under various decay heat removal operations using the DHX.

It was observed that in the long term, a cliff effect occurs, meaning that the apparent steady natural convection is perturbed if a threshold is reached. Instabilities and flow rate oscillations from positive to negative values in the primary loop are observed after a period of smooth natural circulation. The unstable behaviour results from the competition between IHX and DHX cooling, the latter leading to an increase in thermal stratification in the hot pool. This paper aims to analyse this phenomenon, bring a comprehensive criterion for the onset of instable behaviours and give some general guidelines to avoid such effects for accidental transient management.

The evaluation of reactivity feedback in sodium-cooled fast reactors owing to core deformation during the power increase needs a comprehensive understanding of the interactions among neutronics, thermal-hydraulics, and structural mechanics in the core. However, conventional reactor core design evaluation methods often lack accuracy due to oversimplifications in modeling. To deal with this, JAEA has developed an evaluation method that couples several analysis codes implementing detailed models of these phenomena. In the neutronics calculation, core deformation reactivity is based on the first-order perturbation theory and GEM reactivity is determined using a function of core flow rate based on Monte Carlo calculation results of the reactivity. Other reactivities due to the Doppler effect, density reductions of fuel, cladding, coolant, and wrapper tube, and the axial thermal expansion of control rods are calculated by multiplying their temperature increases by their respective reactivity coefficients. The thermal-hydraulics inside fuel assemblies and inter-wrapper regions between neighboring assemblies are modeled as flow networks. The deformation of assemblies is modeled by FEM beam elements. These codes are coupled and synchronized depending on the time scale of each physical phenomenon’s variation to effectively simulate core transients. In this study, the evaluation method was validated by FFTF LOFWOS Test #13 analysis. Comparison between the analyses and test results revealed that the analyses had uncertainties concerning the inclination of the assembly on the core support plate, pad stiffness, and the temperature flattening effect of inter-wrapper flow, which influence deformation reactivity. These uncertainties need further investigation for accurate analysis.

Fuel failures during normal operations and transient scenarios involve rather high uncertainty due to various factors, such as defects in manufacturing, operating conditions, cladding dose, fuel cladding chemical interaction, fuel cladding mechanical interaction, plenum pressurization and cladding thermal creep. While the failure of a single fuel pin poses minimal risk by itself, the potential for pin-to-pin failure propagation (or decrease in failure margin of the neighboring pins) may exist within a fuel bundle. Numerous studies have explored the potential for cascading pin failure, but only in-pile tests with live fuel have created the sudden rupture and rapid fission gas release resulting from cladding failure. A unique burst technique has been developed at Oregon State University to replicate the depressurization of fission gas during fuel failure. This was achieved by laser-welding thin stainless-steel film, laser-etched to create defects, onto partially voided surrogate fuel pins. These pre-defected surrogate fuel pins were then inserted into a 19-pin quartz stainless-steel surrogate fuel bundle, that allowed for the visualization of gas release within typical liquid metal fast reactor (LMFR) geometry and dimensions using a matching index of refraction technique. Failures within the bundle were tested at various burst pressures, coolant flow rates, and breach sizes to characterize the gas release component of fuel failure in a controlled separate effects test. The data from these experiments will inform the design and experimental parameters for future tests in sodium flow loop and contribute to validation of models for unprotected transient events, which currently lack corresponding experimental data.

Japan Atomic Energy Agency has started developing analytical technologies to simulate disrupted core of metal-fuel sodium-cooled fast reactors. This paper reports the development of physical models implemented into the SIMMER code for metal-fuel fast reactor simulations and results of in-pile experiment analysis as a code validation.

To apply the SIMMER code to the metal-fuel fast reactor, priority is given to implementation of two feasible models to represent phenomena specific to a fuel damage accident in the reactor. One of the feasible models is a eutectic formation with a contact of fuel and steel, and the other is an in-pin behavior of molten fuel slug with low melting point. The eutectic formation is treated both in the pin and after pin failure. Furthermore, a cladding failure due to a cladding thinning by the eutectic formation and a molten fuel discharge through the cladding failure can be represented by combining the two models.

To validate the implemented models, this study performed an analysis of the TREAT experiment. The calculation shows that the eutectic formation thins cladding at a top of fuel slug and the cladding failure occurs. The molten fuel in the pin is discharged from the cladding failure to a coolant flow channel. The new models improve the pin failure and a formation of blockage by broken pin and a eutectic material which was observed when not using the models.

The S-Allegro is a state-of-the-art, electrically heated, downscaled Integral Test Facility (ITF) of the ALLEGRO Gas-Cooled Fast Reactor (GFR) demonstrator, operated by CVR in Pilsen, Czech Republic. The facility is designed to investigate operational states and transients of the ALLEGRO GFR demonstrator and to serve as a platform for testing innovative systems and components for the gas-cooled reactor technology. Additionally, it aims to generate experimental data for the validation and verification of thermal-hydraulic (TH) codes and models used in further ALLEGRO research and development.
Two of the most critical components of the S-Allegro ITF are the primary (gas-gas) and secondary (gas-water) heat exchangers. This paper presents a summary of the TH benchmarking activities conducted as part of the Euratom project SafeG. Participants from VUJE, HUN-REN EK, CVR, BME and NCBJ developed TH models of the primary and secondary S-Allegro heat exchangers using the RELAP5-3D, CATHARE2, DYMOLA, ATHLET and APROS codes, and performed a complex steady-state qualification. The outcomes of these activities, along with recommendations for further benchmarking, are discussed in this paper.

ALLEGRO is a demonstrator for the large GFR2400 gas-cooled fast reactor selected by the Generation IV International Forum (GIF). These have been under development in Europe for more than two decades. The primary aims of ALLEGRO are to demonstrate helium technology and to provide some technological background to test the new refractory fuel in a fast-spectrum environment. Two main core configurations are envisaged in ALLEGRO. The first is the so-called driver core, which consists of MOX or UOX fuel with stainless steel cladding. The second is the refractory core aiming to utilise SiC-SiC cladding and carbide fuel. In this study, we carry out thermal-hydraulics calculations for the new refractory core, which was proposed in the SafeG EU project. Two transients are investigated with the CATHARE thermal hydraulics code. First, a 200% break at the hot duct is initiated, which does not lead to loss of coolant but causes a serious core bypass. The second transient describes the evolution of a LOCA transient at the cold duct. The results are compared to the simulations carried out for the former refractory core.

Nuclear power generation has advantages, such as high energy density, reliable power supply, and a reduction in greenhouse gas emissions. However, the potential risk of nuclear accidents requires increased reliability. High Temperature Gas-cooled Reactor (HTGR) is a new generation of reactors that operate at high temperatures above 750°C. This high thermal energy can be used not only for power generation but also for industrial heat applications and hydrogen production. HTGR improves safety with a Reactor Cavity Cooling System (RCCS), which is a fully passive system requiring no external power or coolant. When the active cooling system of the reactor core is off, the RCCS transfers decay heat from the reactor core to the concrete walls of the reactor cavity. In the RCCS, a vertical rectangular riser duct surrounds the reactor vessel at a certain distance, and a chimney connects to the riser duct. The riser duct receives the decay heat from the reactor vessel and the rising air is released to the external atmosphere by natural circulation, maintaining the safe temperature of the reactor. During this process, most of the heat is transferred in the form of radiation. In this study, a preliminary validation of the radiation model comparison for radiative heat transfer analysis of the air-cooled RCCS of MHTGR is performed by Computational Fluid Dynamics (CFD) analysis.

Based on the efficient thermoelectric conversion capability of Stirling engines, the Autonomous Circulation Micro Integrated Nuclear Reactor (ACMIR)is highly integrated and lightweight, making it a favorable candidate for deep space exploration, manned spaceflight, and other projects. However, in demonstrating the applicability and safety of ACMIR across various application scenarios, challenges arise due to the lack of simulation calculation models and modeling methods that account for multi-parameter physical coupling. Therefore, this study considers the heat source structure of the reactor core integrated within the Stirling engine to establish a refined system thermodynamic model. Additionally, it establishes a dynamic model for the pistons, considering the reciprocating motion of the gas-distribution piston and power piston in the Stirling engine. Subsequently, the transient neutron dynamics and the mathematical differential equations for the electromagnetism of the moving-coil linear generator are coupled and solved, completing the multi-physical parameter coupling calculation for the "nuclear-thermal-mechanical-electrical" aspects of ACMIR. By selecting appropriate mathematical algorithms for model solving, preliminary characteristic analysis of ACMIR under different load conditions is conducted. The analysis results indicate that the established simulation model can basically align with the operational states of the space reactor system under different mission conditions. The developed model can serve as a research reference for the next step in system control.

The S-ALLEGRO Integral Test Facility (ITF) is a downscaled version of the ALLEGRO Gas-cooled Fast Reactor (GFR) demonstrator. Benchmarking the measurements conducted on this facility is crucial for the safe and effective development of ALLEGRO. The benchmark activities require participants to create thermal-hydraulic models of the whole S-ALLEGRO system. Since the system consists of several different and innovative components, the modelling approach is to look at the different heat exchangers, blowers, and pipelines and create a standalone model for each. If measured data is available for the separate components, the validation of the standalone models is essential to get reasonable calculations for the whole facility.

The modelling of the heat exchangers is probably the most critical part of the benchmark from the calculations point of view. One of these heat exchangers is a shell and tube type with U-shaped tubes and baffles inside it. It is called the Secondary Heat Exchanger (SHX) in S-ALLEGRO, and it has helium on the tube side and water on the shell side. Having baffles on the shell side can make the waterside flow pattern complex, so special attention has to be paid to its modelling by a 1D code.

In this paper, a special way of modelling such a heat exchanger in the ATHLET code is presented which is supported by standalone measurements.

High Temperature Gas-cooled Reactor (HTGR) with high outlet temperature from 700°C to 800 ~ 1000°C is expected to be widely used for heat supply, hydrogen production, steelmaking, seawater desalination, thermal recovery of heavy oil, coal liquefaction and gasification. The 10 MW High Temperature gas-cooled test Reactor (HTR-10), with outlet temperature of 700°C, had been constructed and operated in China as a pilot plant to demonstrate the inherent safety features of the modular HTGR. Supported by Chinese National S&T Major Project and National Key R&D Program of China, some research on HTGR technology with much higher outlet temperature is carried out.

This paper presents results obtained for two Beyond Design Basis Accidents: (1) control rod withdrawal ATWS and (2) control rod withdrawal ATWS combined with DLOFC. Within a cooperation between the Nuclear Research Group (NRG) of Netherland and Institute of Nuclear and new Energy Technology (INET), Tsinghua University of China analysis was performed with two different codes, TINTE code, a thermal-hydraulic design and accident analysis tool for the Pebble-bed High Temperature Gas-cooled Reactor (HTGR), and the SPECTRA code, a thermal-hydraulic analysis code developed at the NRG.

The performed calculations showed that the fuel temperature will stay below the acceptable limits set for the DBA (1620ºC) during the accidents. The results show feasibility to increase the outlet helium temperature of the HTR-10 to 950°C.

This study presents an experimental investigation on compact diffusion-bonded steam generators, namely Printed Circuit Steam Generator (PCSG) designed for high-temperature gas reactors. The thermal performance of the PCSG was evaluated utilizing the High-Temperature Helium Experimental Facility at the University of Michigan, which enables the characterization of single-phase and two-phase flow heat transfer in the PCSG’s mini-zigzag channels. In this study, two PCSGs with different flow channel design, i.e, straight channels and zigzag channels, were tested under helium-to-water/steam heat transfer setup. The heat transfer characteristics of both the PCSGs were analyzed based on the measured parameters, including the system pressure, mass flow rate, and temperature data. The averaged two-phase heat transfer coefficient inside the cold channels was found to vary with the vapor quality at the cold channel outlet. A sharp drop in the two-phase heat transfer coefficient was observed when the cold channel outlet vapor quality was around 0.5 – 0.6 due to local dry-out of the thin liquid film at the annular flow regime. In addition, the zigzag channel PCSG exhibited enhanced convective boiling heat transfer, with a higher heat transfer coefficient compared to the straight-channel PCSG in the high vapor quality region. However, in the low vapor quality region, significant measurement uncertainties were observed due to the high sensitivity of the evaluated heat transfer coefficient to the helium-side single-phase heat transfer coefficient. The findings from this study provide valuable insights into the design optimization of compact steam generators for next-generation small modular reactors and micro modular reactors.

Given the rising global energy demand, Generation IV lead-cooled fast fission reactors (LFRs) are emerging as a promising technology, offering inherent safety, reliability, and sustainability. In this framework, newcleo is planning to demonstrate the feasibility of building a first-of-a-kind 30 MWe LFR (LFR-AS-30) by the early 2030s.

One of the decay heat removal systems (DHRSs) envisaged for newcleo’s LFR, designed to remove the residual heat generated by the core after a shutdown, is based on the dip cooler architecture: several tens of bayonet tubes, working in parallel, are directly submerged into the reactor’s primary coolant pool. Water, the secondary fluid, flows through each bayonet tube undergoing phase change.

To assess potential instabilities that may occur within the dip cooler DHR, the Dip Cooler Instability (DCI) Test Facility was designed. The facility will be operated to investigate the behavior of two bayonet tubes. The primary focus of the activity has been the computational modeling of the DCI Test Section using thermal-hydraulic system codes. The selected system codes are ATHLET 3.4.1 (2023.2) and RELAP5/MOD3.3. The models comprise two coupled bayonet tubes operating in parallel, with a uniform and constant temperature applied to the outer surface of both risers. To support the facility design phase and test matrix definition, a code-to-code benchmark was performed prior to experimental validation.

The results of the different codes are compared to highlight the level of agreement. The current models will be extended to the entire facility and will be validated against the upcoming experimental campaigns.

Due to its unique safety and economic advantages, the lead bismuth eutectic (LBE) cooled fast reactor has been extensively studied. A multi-application experimental circuit (MATH) for LBE was constructed, and the steady-state and heat transfer characteristics of the circuit were investigated with different heat flux. The fluid temperature distribution in each section of the test section was measured to obtain the convective heat transfer coefficient. The experimental results indicate that the LBE exhibits stable flow characteristics in the heating power range of 13-20 kW. the Pe number remains basically constant across different heating powers, indicating that the flow characteristics are independent of the heating power. Experimental and theoretical analyses demonstrate that for upward flow, the heat transfer coefficient decreases with increasing heat flux, indicating that the buoyancy effect enhances the impairment of heat transfer. Based on the experimental data, a new LBE convective heat transfer correlation is proposed, and its relative error with the experimental data is less than 5%.

Lead-bismuth alloy has the characteristics of high thermal conductivity, low melting point and high boiling pointcite{xing2025comparative}, which enables the lead-bismuth liquid metal-cooled fast reactor to operate at atmospheric pressure and achieve a high core average temperature, and thus it has a unique advantage over the traditional pressurized water reactors in terms of safety and economy, making it a fourth-generation nuclear energy system and has a wide range of application prospects. According to the different uses of lead-bismuth reactors, it is of theoretical value and practical engineering significance to carry out related technical research.

This paper takes the small integrated lead-bismuth reactor as the research object, and establishes the simulation model including the lead-bismuth reactor vessel and the main cooling circuit, electromagnetic pump, helical coil tube type once-through steam generator model and voltage regulator based on the system analysis program NUMAP, and establishes control modelsd for the reactor power, the steam generator, the gas pressurizer, and the electromagnetic pump, based on the operational characteristics of the lead-bismuth reactor. Through simulations under normal operating conditions and accidental conditions such as control drum stoppage, it is demonstrated that the established simulation model accurately reflects the steady state characteristics of the system. The verification of transient lift power is also completed, and the control model effectively regulates the system. This lays a foundation for in-depth research on the operational and control characteristics of the lead-bismuth reactor power unit.

Argonne National Laboratory participates in the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop (NACIE)”. The benchmark project includes three experiments from the NACIE lead-bismuth eutectic (LBE) experimental loop located at the ENEA Brasimone Research Center, in Italy. All benchmark tests are on transition from forced to natural circulation initiated by shut-off of the gas lift-off pump. The main difference between the tests, ADP10, ADP06, and ADP07, is which heater pins are activated (heated) during the tests, meant to approximate partial flow blockage in a fuel assembly.

Argonne’s work with the SAS4A/SASSYS-1 code on the CRP includes system-level and sub-channel simulations. Via comparisons against experimental measurements from the NACIE tests, these benchmark simulations are being performed to expand the validation basis of the SAS4A/SASSYS-1 code. The paper presents progress on NACIE test simulations for simulation of the NACIE benchmark tests with SAS4A/SASSYS-1 code.

All the results obtained so far for NACIE tests and presented in this paper in general show good agreement with the available experimental data. However, in some cases, model modifications were needed to obtain that good level of agreement - those model modifications are also presented in the paper, along with the identification of the remaining differences and approaches for how to resolve them in future work.

Utilizing a fuel assembly cooled by a lead-bismuth alloy within a narrow rectangular channel offers significant benefits for thermal exchange. This design enhances the thermal transfer area within the core, enabling the efficient removal of excess heat. In this study, we focus on understanding the role of surface roughness in narrow rectangular channels for lead-bismuth alloy. By using detailed numerical simulations, we explore how variables like the type, height, and spacing of the roughness affect the flow and heat transfer characteristics of the alloy. Our findings indicate that the channels with a rough interior have a much higher Nusselt number and friction resistance compared to channels with a smooth interior. The disturbance of velocity distribution around the roughness elements significantly affects surface resistance, turbulent mixing, and heat transfer. When the fluid flows over these roughness elements, the fluid behind them is disturbed and forms vortices, which disrupt the flow and heat transfer boundary layers, thus enhancing heat transfer and also increasing flow resistance. These results offer valuable insights for the design of high flux reactor core.

The development of Generation-IV (Gen-IV) reactors is accelerating globally to enhance safety and support diverse applications beyond electricity generation. Among these, the Molten Salt Reactor (MSR) stands out for its use of molten salt as fuel and coolant, enabling high operating temperatures and efficient heat transfer. This design offers inherent safety advantages, such as reduced meltdown risk and passive safety features. In Molten Salt Fast Reactors (MSFRs), the Gaseous Fission Products (GFPs) were removed by helium bubbles. Additionally, the helium bubbles could be used to control reactor reactivity. However, the complex interactions between helium bubbles and molten salt present challenges that traditional computational methods struggle to predict.

Understanding the dynamics of helium bubbles is essential to model these interactions accurately in MSFRs. Also, it could help to improve the efficiency of fission gas removal and the accuracy of the model that describes the physical phenomena in numerical simulation. Despite its importance, helium bubble dynamics have not been thoroughly explored. To address this, a multi-physics framework was implemented using the Volume of Fluid (VOF) method to track gas-liquid interfaces and the PoliMi neutronics model to simulate reactivity changes driven by helium bubbles.

Numerical simulations were conducted to study the impact of helium mass flow rates and injection points on bubble motion, deformation, and distribution. The results enhance our understanding of multi-phase flow dynamics in MSRs and provide critical insights for optimizing reactor performance. Moreover, the findings offer valuable data for AI-based analyses, aiding the design of safer, more efficient reactors.

Fission products in liquid-fueled molten salt reactors are often categorized as soluble salt-seekers, weakly soluble noble gases, and weakly soluble noble metals. Noble metal fission products include Mo, Tc, Nb, Ru, Te, Ag, etc. Based on the operation experience from the Molten Salt Reactor Experiment, these noble metals tend to separate from the salt phase and migrate to the interfaces presented in the reactor system, including heat exchangers, graphite moderator, entrained cover gas bubbles, liquid surface in the pump bowl, etc. The uncontrolled migration and deposition of noble metals negatively impacts the neutronics, radiation protection, and thermal-hydraulics of the reactor. In this study, molecular dynamics (MD) simulation is used to investigate the microscopic behavior of representative noble metal constituents in molten salts. The polarizable ion model is implemented in the LAMMPS code and open-sourced. The code implementation is verified against theoretical results and existing simulation study with CP2K. The model is validated against the experimental density, viscosity, and diffusivities of the base salt. After the verification and validation, the diffusivities of the noble metals in typical fuel salts are simulated, and comparison is made with the Stokes-Einstein correlation. Lastly, preliminary studies on the aggregation of noble metal molecules in molten salts are presented. This phenomenon is important as the formation of critical nucleus of noble metals from aggregation is the first step in the migration of noble metals in liquid-fueled molten salt reactors.

Pronghorn, a thermal-hydraulics code developed within Idaho National Laboratory's Multiphysics Object-Oriented Simulation Environment (MOOSE), has been adapted to model tritium production and transport in the molten salt tritium breeding blankets of fusion reactors. This work highlights recent developments in Pronghorn that enable detailed simulations of two-phase flows, mass transfer between phases, and the volatilization of tritium in molten salt systems—critical aspects for sustainable tritium production, fusion system safety, and tritium management strategies. At the core of Pronghorn's capabilities for this application are its two-phase mixture models, which allow for the simultaneous tracking of liquid and gas phases. These models incorporate mass transfer mechanisms that control tritium migration between the molten salt and gas phases. The integration of Thermochimica, a thermochemical equilibrium solver, provides accurate modeling of tritium volatilization and chemical speciation in molten salts, enabling a comprehensive understanding of tritium behavior under fusion system operating conditions. A case study is explored, focusing on tritium production in a molten salt blanket and its transport to the gas phase. The impact of key factors such as temperature gradients, flow patterns, and salt composition on tritium release and containment is examined. Finally, future development directions are discussed, aimed at further enhancing Pronghorn's predictive capabilities for tritium dynamics in molten salt tritium breeding blanket systems.

The analysis of advanced reactor concepts such as the Molten Salt Reactor (MSR) requires the development of new modelling and simulation tools to deal with the characteristic features brought by the innovative design. One of the peculiar aspects of liquid-fuel reactors such as the MSR is the mobility of fission products (FPs) in the reactor circuit. Some FP species appear in the form of solid precipitates carried by the fuel flow and can deposit on reactor boundaries (e.g., heat exchangers, fuel containment walls), potentially representing design issues related to the degradation of heat exchange performance or radioactive hotspots. The solid FPs tracking is therefore relevant for the prediction of these phenomena. For this problem, both the Eulerian-Eulerian (E-E) and Eulerian-Lagrangian (E-L) approaches can be used, however, while the former can only track a scalar field representing the average concentration of FPs, the latter allows to individually track the behaviour of solid particles inside the reactor domain. Treating the particles as physical bodies instead of scalar fields allows for a proper introduction of the phenomena influencing its behaviour, especially for deposition. For this reason, an E-L based solver is verified against an analytical case. This case was previously developed for the verification of an E-E multiphysics solver developed at Politecnico di Milano. The benchmark case was adapted for an E-L approach in OpenFOAM with the modification of a pre-existing solver. The verification was done by comparing the solid FPs concentration profiles obtained by the CFD simulation and the analytical case.

Insoluble fission products, including noble metals and noble gases, can significantly impact the operation of molten salt reactors. For example, noble metals tend to deposit on structural surfaces, potentially altering local heat transfer capabilities and, in severe cases, clogging narrow tubes like those found in heat exchanger. Meanwhile, noble gases like Xe-135, which has a high neutron absorption cross-section, must be efficiently removed from primary loop to minimize reactivity effects. The removal of these insoluble fission products from the primary loop is typically achieved through a gas sparging process, where the characteristics of the bubbles play a crucial role in determining the efficiency of insoluble fission products mass transfer to circulating bubbles. Research indicates that noble metals form surfactants at the bubble interface, making the bubble interfaces more rigid and thereby decreasing the efficiency of mass transfer. However, the precise impact on fissional products removal due to surfactants has not yet been fully explored. This study addresses this gap by modeling a time-dependent mass transfer coefficient, mimicking the gradual surface contamination process on cover gas bubbles. It further examines how this varying coefficient influences the distribution of noble metals throughout the MSRE loop. This approach enables a quantitative analysis of how surface surfactants affect the efficiency of mass transfer to circulating bubbles. The findings can provide valuable insights into optimizing fresh cover gas injection frequency, ultimately improving the removal of insoluble fission products from MSR primary loop.

The decay product of molybdenum-99 (Mo-99), technetium-99 m (Tc-99m), is a common, short-lived radioisotope used in medical imaging. Several technologies are being investigated as alternative means for producing Mo-99. Online extraction from Molten Salt Reactors (MSRs) through electrochemical deposition is one such technology that is used as the motivation for this paper.

The purpose of this paper is to perform a validation, verification, and uncertainty qualification study on a flow past a cylinder model acting as a simplified model of online Mo-99 deposition in an MSR. One reason for this study is to examine the feasibility of using Reynolds Averaged Navier Stokes (RANS) models to accurately simulate mass transfer on a cylinder in external flow. The RANS results are compared to experimental and Large Eddy Simulation (LES) heat transfer results to determine their accuracy because mass and heat transfer are analogous. Another reason is to determine the mesh complexity needed to produce accurate results. An extensive mesh independence and input uncertainty study is performed on each baseline mesh.

From the validation, verification, and uncertainty quantification study, RANS models are determined to be accurate before the separation angle of the cylinder but overestimate the mass transfer in the wake region. LES is needed to estimate this turbulent recirculation region. A fully complex mesh usually used in flow past cylinder simulations is not needed for mass transfer simulations with RANS models. Simpler meshes are sufficient and reproduce similar results while reducing the computational time.

The most critical task in developing microreactors is investigating the thermal-hydraulics of long-length sodium heat pipes. Although research on sodium heat pipes has increased in recent years, the manufacturing and testing of long (~4 meter) heat pipes remain significant challenges. Given their importance as a key milestone in microreactor development, this paper aims to validate the thermal similarity between sodium and water heat pipes using scaling laws. These laws are applied by matching dimensionless numbers related to pressure distribution, such as pressure drops in the wick and vapor, with the geometry and boundary conditions of the heat pipes determined through a 1D pressure distribution analysis code. The results indicate that the largest discrepancy between sodium and water heat pipes arises from differences in vapor inertia drops due to thermal properties. To verify this similarity, experiments are conducted on 900 mm sodium-water pipes. Based on these experimental results, we aim to predict the thermal distribution of a 4000 mm sodium pipe. Additionally, by conducting experiments with heat pipes of varying lengths, the study seeks to analyze how length impacts heat transfer behavior and to further validate thermal similarity under different conditions. Leveraging these insights, this study will assess the potential application of long (~4 meter) heat pipes in microreactors.

Heat pipes play a crucial role in microreactor cooling systems, valued for their high efficiency and compact design. However, optimizing their performance remains a complex challenge, particularly when considering critical factors such as filling ratio and inclination angle, both of which can significantly influence heat transfer efficiency. These variables can substantially affect heat transfer efficiency. This study addresses the gap in existing numerical investigations of filling ratio effects on heat pipe performance, with a focus on conduction-based models. While most existing codes primarily solve vapor and liquid flow dynamics to evaluate heat pipe performance, these calculations are computationally expensive and highly unstable. In contrast, conduction-based models offer a faster and more efficient alternative but have lacked proper implementation of filling ratio effects. In this work, we develop a new heat pipe performance code that incorporates filling ratio into a twodimensional conduction model. This approach provides a more practical and efficient solution for analyzing heat pipe behavior, making it well-suited for applications where computational resources are limited. The model was tested under both steady-state and startup conditions, with system coupling to computational fluid dynamics (CFD) programs to evaluate its performance. The preliminary validations results show good agreement with the experimental data.

Solid state heat pipe reactors have a broad prospect in the fields of sea, land, air and space. The high temperature heat pipe, as the most critical heat transfer component in a heat pipe reactor, has a high priority on the wick. However, the influence of key parameters such as permeability, porosity, and capillary force of the wick structure on the working fluid distribution inside the heat pipe is difficult to measure experimentally. In this study, firstly, a mechanistic experiment with a wick is used for performance testing, followed by a three-dimensional CFD model of a heat pipe with a wick structure, which can predict the heat transfer characteristics under different steady state conditions. Based on Star-CCM+ numerical simulation software, the effects of fluid flow and convergence behavior within the wick structure on the heat transfer characteristics of the heat pipe were simulated using a combination of a liquid film model and a volume of fluid (VOF) model. Sodium high temperature heat pipe experiments were used to verify the accuracy of the numerical simulations with a maximum error within 10%. The effects of operating angle and wick structure on the heat pipe are investigated, and this study lays the foundation for the design and analysis of the heat transfer characteristics of high-temperature heat pipes.

A Small Modular Reactor (SMR) requires advanced safety features capable of providing long-term passive cooling and maintaining the integrity of the final heat sink. Many designs of a water-cooled SMR employ a water reservoir as the final heat sink, but this water can potentially dry out during an accident due to decay heat. Therefore, in this study, a modular passive-cooling tower equipped with a wickless heat pipe is proposed as the passive safety system to delay the complete depletion of water through air convection. Since the modular passive-cooling tower transfers heat from the final heat sink to the ambient air by air natural circulation, the analysis of natural circulation and the convective heat transfer is crucial to assess its feasibility of the modular passive-cooling tower. An in-house developed code, a nuclear reactor thermal-hydraulic system code, and a CFD program were used in the assessment and their results were compared to each other. Additionally, the system code was used to optimize the system design, which has design to achieve the best heat removal capability. Consequently, this study evaluated structural effects on the overall heat transfer performance and confirmed that the modular passive-cooling tower has significant potential in delaying the depletion of the water in the final heat sink, contributing to the long-term passive cooling capability of the SMR.

This study examines the free piston Stirling engine (FPSE) as a promising candidate for supporting compact, long-lasting, and high-efficiency microreactors suited to remote operation. A 20 kW FPSE was designed and analyzed using Sage software, producing about 19.86 kW at 26.48% thermal efficiency. Key losses, including friction in the regenerator’s wire mesh, conduction and shuttle heat losses, and regenerator performance, were optimized to meet design constraints. Integrating a linear alternator created a free piston Stirling generator (FPSG) that converts mechanical to electrical energy at around 18.99 kW_e and 97.62% conversion efficiency. EMWorks2D, an extension of SOLIDWORKS, helped visualize and refine the permanent magnet configuration, improving the magnetic field path and validating the Sage model. This theoretical investigation supports the viability of FPSEs in microreactor applications. Further improvements could include employing heat flux calculations rather than standard heat exchangers, replacing wire mesh with robust foil in the regenerator, and performing 3D modeling simulations to maximize the engine’s potential.

This study investigates the intermittent boiling phenomenon in sodium heat pipes with wick structures during startup. The experiments focus on the effects of heating power, filling amounts, and capillary wick support structures on temperature oscillations and flow instabilities. A novel stainless steel porous thin-walled tube was designed as a wick support to enhance structural stability. Key findings reveal that intermittent boiling primarily occurs during the continuous flow region expansion phase, characterized by periodic temperature fluctuations. The amplitude and period of oscillations are non-monotonically influenced by heating power, peaking at 800 W with a maximum amplitude of 155.2°C and a cycle of 220 s. Reducing the liquid filling volume decreases oscillation intensity but risks localized dry-out at high heat fluxes. Comparative tests between porous thin-walled tubes and conventional 50-mesh screen supports demonstrate that the former reduces the stable power threshold by 42% by mitigating vapor-liquid interfacial shear stress in the adiabatic section. The study establishes optimal parameters for balancing heat transfer efficiency and operational stability, providing critical insights for the design of high-temperature alkali metal heat pipes in nuclear reactor applications.

To facilitate the deployment of High Temperature Gas-cooled Reactors (HTGRs), modeling and simulation tools that have been validated for such systems are required. The most common methods for HTGR systems analysis are lumped parameter System Thermal Hydraulic (STH) codes that were originally developed and validated for Light Water Reactors (LWRs). The Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD/NEA) is currently administering a benchmark that provides a set of Verification and Validation (V&V) problems and exercises using high quality experimental data from the Oregon State University’s (OSU’s) High Temperature Testing Facility (HTTF), a 1:4 scaled Integrated Effects Test (IET) of the General Atomics’ (GA) MHTGR design.

The OECD/NEA benchmark consists of three separate problems to be analyzed, one of which is the Lower Plenum (LP) mixing exercise. This problem can be tackled in two cases: a code-to-code comparison study with fixed boundary conditions mimicking the full power conditions of the experiment and a code-to-experiment comparison study with best estimate boundary conditions, the former being the focus of current efforts. Previous articles have respectively showcased the time independence study on a coarse mesh and the results of a medium resolution mesh.

The current article presents the grid-independence study using three mesh sizes. All cases use Unsteady-RANS (URANS) solvers employing the Realizable K-Epsilon turbulence model as available in the commercial code Simcenter STAR-CCM+. The results show that although not all considered points converge, the obtained Grid Convergence Index (GCI) is quite low.

MARVEL is a natural-convection-cooled sodium-potassium microreactor that is anticipated to generate 85 kilowatts of thermal energy. It will operate within Idaho National Laboratory’s Transient Reactor Test Facility and is being developed by the DOE Microreactor Program. MARVEL will be used to test microreactor applications, generate operational data, and pave the path for commercial demonstrations.

A thermal-hydraulic computational model of this facility is a valuable tool to study important transients and calculate the safety limits of the micro-reactor design. For this purpose, the authors propose to use a multiscale coupled simulation: SCM for modeling the reactor core and SAM for the reactor’s primary cooling system.

SCM is MOOSE physics module for subchannel analysis, which was designed to model single-phase flows through liquid-metal cooled, wire-wrapped fuel pin sub-assemblies, ordered in a triangular lattice. The SCM code was modified to be able to model MARVEL’s unique geometry.

SAM is a systems analysis module based on the MOOSE framework. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for various advanced reactor types.

The coupling between the two SCM and SAM for MARVEL modeling is done implementing a domain over-lapping approach. The resulting coupled simulation can model transients such as reactor startup/shutdown and provide an intermediate fidelity picture of the temperature field and other variables, in the core. Results for the steady-state simulations are presented in the article as well as flow blockage transient.

Micro Modular Reactors (MMRs) present a promising solution for decentralized power generation, particularly in remote areas. Among the various designs under investigation, Heat Pipe-Cooled MMRs (HP-MMRs) have gathered significant interest. As power conversion unit (PCU), two distinct cycles are being investigated: an open-air recuperated Brayton cycle and a supercritical CO₂ (sCO₂) recuperated Brayton cycle. The goal of this research is to provide a useful comparison at a system level between these two power conversion strategies, offering insights that could inform future design choices for HP-MMRs in off-grid applications.

The thermodynamic modelling and optimization of the two cycles, employed as PCUs for a 5 MW_th Heat Pipe-Cooled MMR, are investigated employing the system code ATHLET. The analysis focuses on the design of key system components, such as the Heat Pipe Heat Exchanger (HPHX), the recuperator, the ultimate heat sink (for the sCO₂ case), and the turbomachinery. Preliminary findings suggest that the air cycle offers operational flexibility, can leverage the maturity of existing technologies from the power and aerospace industries, and does not require an ultimate heat sink. In contrast, the sCO₂ cycle demonstrates advantages in terms of more compact turbomachinery and higher thermal efficiency. Additionally, the study explores potential control strategies and their feasibility for part-load operations, with the aim of enhancing system adaptability under variable load conditions.