Conference Agenda

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Following the TMI accident, the concept of ‘risk’ was introduced to comprehensively evaluate the safety of complex systems in NPPs. Risk means potential losses that may occur in the future due to certain factors and is defined as the probability of an event occurring multiplied by the consequence of the event. Current methodologies such as Probabilistic Safety Assessment (PSA) are used for risk evaluation. However, these methodologies have some limitations, making them inefficient for assessing the risks of individual accident sequences. To address this issue, this study proposes a methodology to develop risk profiles for each accident sequence through source-term analysis. In this study, a source term analysis uncertainty assessment was conducted specifically for core damage accident sequences of Loss of Feedwater (LOFW) and Small Loss of Coolant Accident (SLOCA) based on the OPR-1000 level 1 PSA model. Based on the result, we quantified the consequences for each sequence and developed risk profiles by visualizing the risk through frequency-consequence curves (F-C curve). This approach can efficiently evaluate the risks of each accident sequence efficiently and present them in a clear visualization. These results contribute valuable information to the risk communication process.

As part of ASNR's development of Level 1 and 2 Probabilistic Safety Assessments (PSA), various support studies are conducted for internal events (IE) PSAs and internal and external hazards PSAs (fire, internal and external flooding, internal explosion, seismic, heat wave…) for operating French nuclear power plants and for some other nuclear facilities. Among these, several thermohydraulic studies are performed using tools developed by ASNR such as the SOFIA simulator (Simulator for Observation of Incident and Accident Scenarios) for the Level 1 PSA, and the ASTEC integral code for the Level 2 PSA. Those tools can simulate a wide range of operational conditions, from full power to shutdown states for 900, 1300, and 1450 MWe PWRs, as well as for the EPR. These thermohydraulic simulations play a crucial role in assessing the kinetic and the consequences of accidental scenarios, to determine whether core or fuel damage occurs, to identify the mitigation systems success criteria and to understand when and how the core uncover. They also contribute to the human reliability assessment by providing available time for diagnosis, decision-making and operator actions. Furthermore, these studies allow the examination of uncertainties inherent to key parameters, such as the size and location break in primary circuit, and their impact on the progression of the accident.

The paper presents the status and perspectives of these studies for PWRs or other facilities and introduces some expectations for possible other reactor designs (e.g. Small Modular Reactors - SMRs).

Passive safety systems employing physical processes and phenomena, such as natural circulation, have been widely applied to the contemporary design of Light Water (LW) Small Modular Reactors (SMRs). The demonstration of passive system reliability requires mechanistic analysis of the system performance in all possible accident scenarios. During the assessment, identification of the “failure domains” i.e. the domains of scenario parameters where the passive system fails to fulfil its mission, and associated “failure modes” of the system is challenging due to a wide range of operational conditions that need to be assessed. The brute-force search is computationally impractical due to the high-dimensional nature of the input space and the significant computational cost associated with Full Model (FM) evaluations.

The goal of this work is to demonstrate the feasibility of using advanced search methods, i.e. genetic algorithm, for the identification of the “failure domain” and its boundaries in the multidimensional space of accident scenario parameters. The primary objective is to improve the search efficiency by reducing a Figure of Merit (FOM) defined as the total number of FM evaluations by the number of identified boundary points. Three frameworks are developed, tested and compared on a benchmark case. The method that integrates GA and Adaptive Triangulation Sampling (ATS) demonstrates a good performance.

The Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI) has directed the NEA Expert Group on Small Modular Reactors (EGSMR) to identify and prioritize knowledge gaps where cooperative research would facilitate the safe deployment of small modular reactors (SMRs). EGSMR is executing a pilot project to demonstrate a method to generate research recommendations for advanced technology (AT-), non-water cooled, designs. High-temperature Gas Cooled Reactors (HTGRs) were selected to pilot this process; however, it is meant to be generally applicable with future application to other AT-SMR technologies. The identification and prioritization of phenomenological knowledge gaps has been built into a procedure to be completed by a task team in coordination with subject matter experts and NEA Working Groups. Phenomena Identification and Ranking Tables (PIRTs) are tools used to identify phenomena important to reactor safety by numerical ranking of importance and knowledge level. In the current work, PIRT information is collected by the task team and sorted into phenomenological groupings before being ranked based on PIRT knowledge gaps, safety significance and suitability for international collaborative experimental research. Consulting with subject matter experts, the task team will refine the prioritization list into a final set of high priority research subjects. In a later phase of the AT-SMR pilot project, these subjects will be linked to experimental facilities and compiled into a final set of detailed research activity recommendations.

The interaction between cooling fluid and solid structures (rods, tubes) in nuclear power plants leads to flow-induced vibrations (FIV). These may cause material fatigue, fretting wear, and in worst case, loss of component integrity. The consequence of this might be high standstill costs due to longer or unplanned outages or safety issues like Steam Generator Tube Rupture accidents. Within the European GO-VIKING (Gathering expertise On Vibration ImpaKt In Nuclear power Generation) project, experimental and numerical efforts are performed to improve the understanding and analysis of FIV in nuclear power plants, as well as to develop and validate advanced numerical approaches for their prediction. Within the fifth work package of the project, activities to analyze the tube vibration behavior in the CEA’s AMOVI experiment were carried out at GRS. AMOVI deals with FIV occurring in tube configurations, exposed to a cross-flow. In this work, the FIV are investigated with a fluid-structure interaction (FSI) approach, based on a structural reduced-order model. Such models are of particular interest due to the excessive computational time necessary for the FIV evaluations. The reduced-order model for the structural domain MOR was coupled to ANSYS CFX for the FSI calculations presented in this paper. This FSI approach was validated against AMOVI data for a single flexible tube positioned in the center of a bundle of rigid tubes.

Fluid-induced vibrations (FIV) are a major cause of component failure in nuclear reactors and are of significant concern when extending operating reactor lifespans. The Go-Viking project aims to ensure that the licensed operating lifetime of aging nuclear reactors in Europe can be safely extended by improving understanding of FIV phenomena within steam generator tube bundles. To achieve this, the GOKSTAD experimental tube bundle has been created to collect FIV data for an inline cross-flow tube bundle operating at a higher Reynolds number than previously documented in the literature. In this presented work, we present results from high-resolution direct numerical simulations (DNS) of the GOKSTAD bundle for comparison with experimental data. Due to the computational cost of DNS, a reduced domain of the GOKSTAD bundle is used in these simulations, consisting of three rows with seven columns (five regular columns and two half-tube columns). The mass flow rate within the tube bundle is 15 m³/s, and the pitch-to-diameter ratio within the bundle is 1.44. The DNS is done using the Department of Energy code, NekRS, while the structural responses are simulated using the Multiphysics Object-Oriented Simulation Environment (MOOSE). Results include velocity fields, pressure fields and displacement data of the center tube, allowing direct comparison with measurements collected within the GOKSTAD bundle. The DNS FIV model generated from these results will support creation of fast-running FIV tools, including reduced-order models.

Key Nuclear Steam Supply System (NSSS) components, such as fuel rods and steam generator tubes, are highly susceptible to Flow-Induced Vibrations (FIV) as a result of the turbulent coolant flow. This can cause known failures of these components, such as Grid-To-Rod-Fretting (GTRF) wear and Steam-Generator Tube Rupture (SGTR), possibly leading to costly reactor outages. Historically, analytical and semi-empirical approaches were used to assess the impact of FIV on the components’ structural integrity. However, these are generally only able to give an order of magnitude estimate of the structure’s displacement. With the increase in computational power though, Fluid-Structure Interaction (FSI) simulations, two-way coupling detailed Computational Fluid Dynamics (CFD) and Computational Structural Mechanics (CSM) codes, are being used more and more. Such FSI simulations are able to give increasingly better predictions, matching reference experimental data quite well, in particular in terms of vibration frequency and displacement amplitude. These simulations generally only consider cylinders undergoing relatively small displacements, avoiding contact. To capture GTRF or SGTR resulting from FIV, generally larger displacements are needed, along with contact between neighboring cylinders or between a cylinder and surrounding fixed structural components. The current work shows results of an initial investigation of performing FIV simulations involving structural contact. It considers a cylinder placed inside a channel and subjected to turbulent cross-flow. Different numerical and modeling techniques have been used to try to successfully resolve the large displacements and structural contact. These are presented, along with results of a first FIV simulation involving contact between the cylinder and the channel walls.

Understanding the behavior of flow-induced vibrations (FIV) is crucial for maintaining the safety of steam generators in nuclear power plants. If left unaddressed, vibrations can lead to tube wear, fatigue, and even failure, which can have profound safety consequences. In U-tube designs where two-phase cross-flow dominates, vibration-related issues are further exacerbated. Fluid-elastic instability is the primary mechanism underlying flow-induced vibration that may damage steam generator tubes. Although fluid-elastic instability in single-phase cross-flow has been extensively studied, its behavior in two-phase flows is less understood. The Horizon Europe project GO-VIKING addresses these challenges through experimental facilities designed to study two-phase cross-flow-induced vibrations. These setups focus on the fluid-structure interaction (FSI) between tube bundles and air-water cross-flows.

This paper presents numerical simulations of FIV in two-phase cross-flows. Simulations cover two-phase flows over single tubes and tube bundles. On the fluid side, we use a two-fluid model coupled with a population balance model to account for bubble poly-dispersity, coalescence, and break-up. The structure motion is modeled using a six-degree-of-freedom rigid body motion solver. The numerical results are validated against experimental data on bubble size distributions, void fractions, fluid-structure forces, and displacement spectra. The findings of this work advance the understanding of two-phase FIV, providing insights critical to the safe and reliable performance of steam generators.

In this work, the vibration characteristics of a propeller rotating in lead-bismuth eutectic (LBE) are studied using fluid-structure interaction (FSI) simulations, which are developed in parallel with an experiment performed at the Belgian Nuclear Research Center (SCK CEN). Both the simulations and the experiment are part of a larger campaign to develop a methodology for characterizing the vibrations in primary pumps of nuclear reactors using heavy liquid metal coolants. These coolants are of interest due to their use in Generation IV nuclear facilities such as MYRRHA , using LBE as a coolant, and LFR, using lead. The high density of this liquid can significantly alter the vibration characteristics compared to when used in air and water, and introduce mode coupling, a phenomenon that is not yet sufficiently understood in the context of heavy liquid metals. The simulations allow for an extensive analysis of the different vibration modes. The investigated propeller consists of three symmetrical blades and is operated at different rotational speeds. First, the structural eigenmodes are calculated in vacuum, using the finite element method (FEM). Afterwards, the fluid and propeller are combined in a two-way coupled FSI simulation. For each mode a particular excitation force is applied to the structure to facilitate the extraction of vibration characteristics by analyzing the free response of the system. The result is the eigenfrequency and damping ratio of that mode in LBE. The results show that this methodology allows for an accurate prediction of the measured vibration response in the test setup.

To improve the long-term operation of NPP, dedicated tools are needed to understand and predict the interaction between cooling fluid and solid structures that may lead to flow-induced vibrations. The paper focuses on a validation test case representative of a steam generator configuration.

In the GO-VIKING project, supported by Horizon Europe research and innovation funding, a new water loop named GOKSTAD has been designed and constructed at the von Karman Institute to characterize a single-phase flow field and study the fluid-structure interaction inside a 7*7 square lattice in cross-flow configuration. The facility is designed to operate up to a remarkable gap Reynolds number of 90,000; a significant leap beyond what is currently available in literature.

The paper will first present the facility's flow field characterization based on PIV measurements at the inlet of the test section, between the cylindrical rods of the lattice, and at the outlet of the bundle. The second part will present the mechanical design and characterization of the moving rigid cylindrical tube developed for the Fluid-Structure-Interaction study. The configuration studied consists of two vibrating tubes inline or side by side inside the square lattice, with a vibration of roughly maximum of 10% of the cylinder diameter.

The results presented are precious data for the numerical team in charge of fluid-structure interaction studies, providing high-resolution boundary conditions and flow field data. The final objective is to have medium-resolution numerical tools to assess structural vibrations in Steam Generators under single-phase cross-flow conditions.

Heat pipe reactors have broad application prospects in deep space power, military bases, marine power and so on. For system level analysis of heat pipe cooled microreactors(HP MicroRx), the coupling between neutronics analysis and thermal-hydraulics analysis was usually solved iteratively. This study aims at establishing a multi-physics coupled simulation framework for performing transient safety analysis of HP MicroRx. This simulation framework is based on the system analysis software RETA. RETA is a multi-physics solver framework for advanced nuclear reactors based on C++ and object-oriented design methods. This work includes the following subtasks: 1) developing the heat pipe modeling capability with a fully-implicit solution method; 2) deforming the geometry of the reactor core so that it’s convenient for the finite volume method(FVM); 3) using the Preconditioned Jacobian-Free Newton-Krylov(PJFNK) method to uniformly solve the heat pipe reactor. In the modeling of heat pipes, a one-dimensional compressible flow model was used to model the vapor core, a two-dimensional axisymmetric heat conduction model was used to model the heat pipe wall and wick region, and the heat pipe wick and vapor core were coupled through a conjugate heat transfer interface. The KRUSTY reactor was selected and analyzed using the above simulation framework. Simulation results match the experimental data produced by Los Almos National Lab well. To conclude, this work provides an accurate and reliable tool for safety analysis of heat pipe microreactors.

High-parameter pressurized water reactors (HP-PWRs) with plate-type fuel operate at a higher power density than conventional pressurized water reactors (PWRs). This is accompanied by higher void fractions and the potential presence of slug flow, which can significantly affect reactor neutronic behavior. However, most neutronics and thermohydraulics analyses for PWRs rely on subchannel codes, and the impact of subchannel homogenization remains uncertain for HP-PWRs. This study models slug flow and bubbly flow in both the XY and XZ planes to investigate the effects of void distribution resolution on neutronic behavior using RMC. For slug flow, subchannel homogenization results in a noticeable overestimation of keff in the XY plane. The maximum relative power deviation (MRPD) between the homogenized and reference schemes reaches 3.70% in the XY plane and 5.38% in the XZ plane. MRPD increases with increasing overall void fraction and gas slug void fraction, as well as decreasing gas slug width and length, while it shows limited sensitivity to variations in small bubble radius in slug flow. For bubbly flow, although void distribution resolution has only a marginal influence on keff, its impact on power distribution is non-negligible—especially as the bubble radius increases, the void distribution becomes more non-uniform, and the overall void fraction rises. The MRPD between the homogenized and reference schemes exceeds 2%. These findings highlight the potential inaccuracies introduced by subchannel homogenization in high-void, non-uniform flow environments. Fine-resolution void modeling is essential for accurate N/TH coupling in HP-PWRs, particularly in reactor safety analysis.

To analyze the startup process and load tracking operation characteristics of a heat pipe reactor with Brayton cycle for thermoelectric conversion, this study employs a hierarchical component model to establish simulation software for the heat pipe reactor system. The study analyzes the impact of key parameters in the Brayton cycle on power generation efficiency. By utilizing the control method of the PID model, the simulation of the heat pipe reactor startup is conducted through the control of Brayton’s rotor speed. The control of the rotor speed under normal operating conditions is achieved by controlling the filling amount of the secondary loop working fluid, thereby exploring the inherent safety characteristics and load tracking operation characteristics of the heat pipe reactor under controlled and uncontrolled rotor speed conditions. The results show that pressure ratio, degree of superheat, and the temperature at the inlet and outlet of the unit significantly affect power generation efficiency; the PID control model can simulate the startup process of the heat pipe reactor, and the rotor speed can be well controlled; compared to uncontrolled rotor speed, the controllable rotor speed results in smaller changes in power and thermal parameters such as fuel temperature during the system’s load increase and decrease processes, which is more conducive to the safety of the reactor core.

High-temperature heat pipes are promising devices for advanced microreactor technologies in terrestrial and space applications. Understanding their performance and safety characteristics is critical to the successful deployment of heat pipe microreactor systems. One key safety consideration influencing the design and operational limitations of heat pipes is the occurrence of dryout in the liquid-wick region. This study investigates the effects of temporary dryout conditions induced by pulsed heat inputs to the evaporator that exceed the capillary limitation. Using water as the working fluid, experiments were conducted to examine the transient response of the heat pipe’s external wall temperatures, internal liquid and vapor temperatures, and vapor pressure under pulsed heat input conditions. Pulse lengths were varied to control the duration and severity of the pulsed dryout conditions and study rewetting and the long-term effects on heat transfer performance. Spatial temperature profiles during transients were obtained using an optical fiber temperature sensor in the vapor core. Thermal resistance and hysteresis were evaluated under steady state conditions before and after pulses to assess their impact on overall heat pipe performance. This study provides valuable insights into the internal two-phase flow behavior during dryout and rewetting of the wick. The experimental data set can be used to benchmark numerical codes and validate computational models. Future work will investigate the effect of pulsed dryout conditions with alternative wick designs, varying filling ratios, and liquid metal high-temperature heat pipes to optimize their design and enhance resilience.

Advanced reactor designs will use sodium heat pipes as the primary means of heat transfer from the core block to the heat exchanger system. Such devices provide an efficient and reliable method for transferring heat over a small temperature gradient and at near-atmospheric pressures. However, robust experimental data is needed to better characterize these devices and provide validation metrics for the Sockeye simulation code. To meet these needs, several heat pipes will be manufactured and tested at high powers (~10 kW) to explore manufacturing repeatability, test operating limits, and measure the properties of the working fluid. This work summarizes the heat pipe design and optimization process used to determine the dimensions of the heat pipes that will be manufactured. Analytical expressions from a variety of sources were used to calculate a theoretical ideal design to meet multiple experimental goals. The wick geometry and properties were tuned to potentially encounter four power limits over the range of operation supported by the experimental facilities. To employ the analytical expressions, a simple yet novel averaging scheme was proposed to account for an annular gap surrounding the wick structure. This averaging scheme was applied to limiting experiments in the literature to evaluate its accuracy. Finally, numerical and analytical methods were applied to evaluate the heat pipe operating conditions to ensure the experimental facilities will be able to test the power limits.

High-temperature heat pipes are critical components in the core of solid-state reactor heat pipe cooled reactors, serving as the exclusive and essential means of heat transfer from the core to the energy conversion system. The Geyser Boiling Phenomena (GBP) of high-temperature heat pipes has a significant impact on the safety and stability of solid reactors. This investigation encompasses the design, fabrication, and testing of sodium heat pipes with varying filling ratios, ranging from 33.3% to 100.1%. An extensive array of experimental studies has been carried out to evaluate the heat transfer properties of these sodium heat pipes under diverse conditions, including different heat transfer rates and inclination angles. The results indicate that the design parameters and operational settings, such as filling ratio, heat transfer rate, and inclination angle, significantly affect the GBP of high-temperature heat pipes. This research combines experimental data with relevant theoretical analysis to establish a semi-empirical relationship for predicting the temperature fluctuation period caused by the GBP of high-temperature heat pipes. Furthermore, based on the improved network thermal resistance model, a GBP analysis model is proposed, providing valuable reference for the design and engineering application of heat pipe-cooled reactors.

NEA has recently organized several large events (2023 Joint Nuclear Safety Research Projects event week, 2024 FRAME workshop) to discuss with its members the benefits of NEA collaborative nuclear safety projects, with, in 2025, close to 60 projects completed and more than 60 years of nuclear safety research. Following these events, the NEA has undertaken the development of a high-level nuclear safety research roadmap to give directions for future nuclear safety research. The roadmap reflects regulators key messages to policymakers, highlighting key nuclear safety research capabilities needs and challenges, and at the same time provides deeper technical insights and research directions in major nuclear safety technical areas to help project operators developing project proposals addressing priority nuclear safety issues. A key recommendation formulated through these initiatives is that NEA should develop a framework for securing and organizing resources for advanced thermalhydraulic experiments for safety assessment of advanced reactor designs including designs relying on passive safety systems and small modular reactors, with scaled experimental infrastructures able to generate high-quality data for the development and validation of state-of-the-art codes used in thermalhydraulic analyses. The framework should also include transverse tasks related to data preservation and knowledge transfer. Relevant roadmap insights and the development status of the new framework will be presented. A companion paper will provide main insights gained from recently concluded NEA projects in the thermalhydraulic area.

The Working Group on the Analysis and Management of Accidents (WGAMA) addresses OECD Nuclear Energy Agency (NEA) activities related to potential design-basis accident (DBA) and beyond design-basis accident (BDBA) in nuclear reactors and related technologies. The group addresses safety issues of existing nuclear reactors and related technologies as well as emerging challenges on evolutionary and innovative reactor designs and nuclear technologies, including Small Modular Reactors (SMRs). For these purposes, the WGAMA has coordinated workshops, technical publications and research activities in the fields of thermal-hydraulics (T/H), computational fluid dynamics (CFD) and severe accidents (SAs) to improve knowledge of accidents and of the confidence of the scientific calculation tools used in safety studies, namely safety analysis codes or tools.

The paper aims to review and summarize the recent activities and outcomes such as the ongoing effort on the extension and update of the CSNI Code Validation Matric (CCVM) and recently completed activities on the applicability of uncertainty quantification methodologies to CFD in the context of nuclear reactor safety studies, design extension condition without significant fuel degradation (DEC-A), reactor pressure vessel integrity assessment for in-vessel retention and the state-of-art-report on behavior of combustible gases in severe accidents.

The Experimental Thermal Hydraulics for Analysis, Research and Innovations in Nuclear Safety (called ETHARINUS) project developed in the frame of the OECD Nuclear Energy Agency (NEA), serves the objectives of investigating complex thermal-hydraulics phenomena. Within this project, issues like the performance of passive heat removal systems and Design Extension Conditions (DEC) scenarios were investigated. The simulation of such events is of high importance to ensure relevant understanding of key thermal-hydraulic phenomena, to perform adequate safety analysis, to assess the efficiency of the adopted accident management procedures and to optimize operator training.

ETHARINUS project is highly relevant for the improvement and validation of thermal-hydraulic safety codes and their use, to maintain a high level of competence and expertise in the field of system thermal hydraulics. It is also a way to gather the actors within the area (safety authorities, operators, experimental facilities operators, university, R&D institutes, etc.) to develop knowledge and common culture about key safety issues, for the operating fleet and for innovative designs.

The objective of this paper is to describe the opportunities of the thermal-hydraulic research facilities employed for the activities, to briefly outline the tests programme, and finally to highlight the key safety issues and safety relevance of these tests programme. Such programmes are conducted in an international context, to share the costs and the expertise, and to promote quicker and deeper international consensus on safety issues. Recommendations are finally proposed regarding how to address the loss of critical research infrastructure (i.e. facilities, capabilities and expertise).

For several years, the OECD Nuclear Energy Agency (NEA) has conducted extensive experimental research through joint projects with broad participation from member countries. These collaborations enable shared costs and expertise, accelerating the global consensus on critical nuclear safety issues.

This paper presents key achievements from recent NEA joint projects on thermal-hydraulics. It outlines the capabilities of the research facilities involved, the critical safety issues addressed, the relevance of the test programs and related analytical activities. These projects aim to investigate phenomena where safety knowledge is insufficient, providing qualified data to develop and validate thermal-hydraulics computer codes used for nuclear safety assessment.

A major product is the experimental data itself, a priority for NEA members. Such data are then used for comparative code assessment to identify strengths and weaknesses of the codes.

Additionally, the paper highlights the importance of international cooperation in preserving unique experimental infrastructure, addressing the challenges of the closure of unique research facilities, and fostering the preservation of expertise while advancing new knowledge. A companion paper will offer insights into the new NEA Framework for Advanced Thermal-hydraulic Experiments for Nuclear Analysis and safety application (ATHENA).

The main outcomes of the following projects will be discussed:

• Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS)
• Experimental Thermal Hydraulics for Analysis, Research, and Innovations in Nuclear Safety (ETHARINUS)
• THAI Experiments on Mitigation Measures and Source Term Issues for Enhanced Severe Accident Management (THEMIS)
• PANDA Experiments Addressing Complex Safety Challenges for Current Water Reactors and Small Modular Reactor Projects

Heat pipes are highly efficient self-contained two-phase passive cooling devices. They are used in a wide range of applications and have recently been investigated as cooling systems for new Micro Modular Reactor (MMR) concepts. The Alkali Metal Heat Pipe Assembly Testing (AHPAT) rig in the single heat pipe configuration has been used in the High Temperature Fuel Channel (HTFC) laboratory of the Canadian Nuclear Laboratories (CNL) to investigate the behaviour of sodium heat pipes near their operational limits. Power is delivered to the AHPAT rig through a heating bank attached to the evaporator of the heat pipe to simulate heat provided by a reactor core. Power output is measured using a gas-cooled stainless-steel block attached to the condenser of the heat pipe. After reaching steady state near the operational limits of the heat pipe, the heaters were pulsed to enable the characterization of transient behaviour. The results of these tests show the temperature distribution of the heat pipe at steady state and near its operational limit. Pulsing the power shows its effect on the temperature distribution as well as the recovery behavior and return to steady state once pulsed heating has stopped. The results of this work will be used for the development and benchmarking of numerical codes that simulate the behaviour of alkali metal heat pipes.

Heat pipe-cooled Micro Modular Reactors (HP-MMR) are mobile systems with a low power output of below 10 MW_el. Terrestrial and space applications are envisaged such as supplying energy to a remote settlement or to a space probe. A heat pipe is a passively working, two-phase heat transfer device that exploits phase change and capillary pumping of the liquid in a wick for the efficient and reliable cooling, e.g. of a reactor core. The high‑temperature heat pipes integrated in HP-MMRs are typically filled with an alkali metal such as sodium or potassium. To enable the safety analysis of a HP-MMR, the thermal hydraulic system code AC²/ATHLET is currently in development for the simulation of high-temperature heat pipes within the MISHA project. A potassium material property package has recently been implemented for the purpose of heat pipe simulations, so that the latest code version provides the fluid properties of sodium and potassium. In addition, many relevant phenomena occurring in a heat pipe have been modelled such as capillary pumping, phase change, radial heat transfer through the wick, friction in each of the phases, and pooling. A verification case will be presented and discussed. The future validation of the module is planed based on upcoming experiments with potassium heat pipes at the IKE Stuttgart which cooperates within the MISHA project.

Heat pipes are essential components in space nuclear reactors which play a key role in facilitating deep space exploration. The temperature difference between the evaporator and condenser, along with the heat transfer limit, are critical performance metrics that govern the thermal efficiency and operational capacity of heat pipes. This study presents an optimization framework that integrates the Genetic Algorithm (GA) with COMSOL Multiphysics simulations to minimize the heat pipe temperature difference while maximizing its heat transfer limit. Key design parameters, including wick thickness, porosity, and vapour core diameter, are systematically optimized using GA to enhance overall thermal performance. Simulation results demonstrate the varying influence of these parameters on heat pipe efficiency, providing valuable insights for optimizing the design and operation of heat pipes in space reactor applications.

The simulation of heat pipe cooled microreactors is a significant challenge, involving tight coupling between specialized codes and solvers. A wide array of governing equations, discretization schemes, and numerical methods may be employed in modelling the involved physics at different degrees of resolution. A flexible and high-performance coupling framework is needed to incorporate such a variety of components in a sustainable way. To this end, this work seeks to assess the usability and performance of the preCICE coupling library for microreactor simulations, with an eye towards nuclear electric propulsion systems. Until now, the use of preCICE in the field of nuclear energy has been limited to surface coupling or “high-low” applications, since the mesh mapping capabilities needed for overlapping 3-D domains have only recently become available in the library. At present, these and other attractive features of preCICE are leveraged, including pre-existing “code adapters” which facilitate the coupling of simulation codes that employ popular PDE libraries such as OpenFOAM, deal.II, and FEniCS. The adapter for OpenFOAM is modified to allow the transfer of arbitrary scalar fields; thereby preCICE is used to couple a custom heat conduction solver with neutron diffusion, as well as point-kinetics. A simplified version of the KRUSTY reactor is modelled under steady-state and transient conditions. The coupling scheme is compared with a standalone-OpenFOAM approach, with good agreement observed. Finally, the viability of the preCICE-based framework for more advanced simulations of space nuclear reactors is discussed.

GRS cooperates with the University of Stuttgart in the MISHA project to establish a calculation chain for innovative MMR designs. Simulation of these new designs comes with unique challenges like rotatable control drums with absorber crescents, solid monolithic cores, and heat pipe cooling. To validate the coupled system of the GRS-codes ATHLET and FENNECS for such simulations, reference calculations based on the Special Purpose Reactor design by the Los Alamos National Laboratory for a heat pipe cooled fast micro reactor will be performed. This design was chosen as reference because some thermal and neutronic data as well as specific reactor parameters are publicly available.

To this end, we present the results of Monte Carlo simulations of the core with Serpent for different absorber configurations, performed to obtain a macroscopic cross-section library and data on delayed neutrons. Core reactivities agree with published values and the operational state is reached with a similar control drum configuration.

Utilizing data from the Serpent results, a first model of the core was created in the neutron diffusion and SP3 code FENNECS and initial stand-alone calculations were performed. Additionally, we show the results for normal operational state with the thermal-hydraulics code ATHLET and point kinetics utilizing the power distributions and delayed neutron data from Serpent. This model adequately reproduces the limited amount of publicly available thermal-hydraulic data for the reference design. In the future, models in both GRS codes will be further refined and eventually coupled to simulate the reactor during operation and in transient conditions.

Heat pipe microreactors provide a unique opportunity for decarbonization as a source of industrial process heat. For example, in the cement production industry, the carbon-heavy kiln flue gas used to decompose calcium carbonate feed meal might be partially or fully replaced by air heated by a heat pipe microreactor. To evaluate the feasibility of such a change, an experiment is being planned to link a single representative heat pipe to a lab scale fluidized bed calcination reactor. The present work summarizes the experimental system design process used in the facility planning. Analytical and empirical correlations from the fluidized bed literature were employed to calculate the air flow rate, temperature, and pressure needed to operate the calcination reactor. This information was then used to inform the design of the rest of the system, which will function as a heated wind tunnel. Special attention was paid to the filtering which occurs directly upstream of the calcination reactor to ensure a predicable velocity profile entering the test section. CFD methods were applied to investigate the calcium carbonate particle distribution in the calcination reactor considering the flow preconditioning under different Reynolds numbers for the determined system design.

In the era of carbon neutrality, low-carbon hydrogen production technology is emerging as a source of hydrogen energy that can replace fossil fuels. As one of the low-carbon hydrogen production technologies, the High Temperature Steam Electrolysis (HTSE) methods that produce hydrogen through electrolysis of high-temperature steam are gaining attention. In this regard, the Korea Atomic Energy Research Institute (KAERI) has conducted an integral test for hydrogen production by coupling a helium loop simulating a high-temperature gas-cooled reactor with a high-temperature steam/air supply system and an HTSE system including two Solid Oxide Electrolyte Cells (SOEC) for hydrogen production. The heat provided by the helium loop is utilized to produce steam and air suitable for HTSE through a steam generator and multi-stream heat exchanger in a high-temperature steam/air production system. These are provided to the SOEC stacks within the HTSE system to produce hydrogen. The integral test results confirmed that the helium loop and the high-temperature steam/air supply system could be coupled to reliably supply the steam and air with suitable temperature, pressure, and flow conditions for the SOEC stacks. In addition, the results demonstrated that two 3kw SOEC stacks at 700℃ with 80A of current produced 4.3kg/day of hydrogen.

Using a Light Water Reactor (LWR) to produce steam for process heating is a topic of rising interest in the industry. LWR process steam has been successfully used in district heating and in petrochemical processes operating at lower pressures and temperatures. However, to make a significant impact on decarbonizing petrochemical plants, higher pressures and temperatures would be advantageous. Previous studies have suggested that high temperature gas reactors would be best suited for this application. However, a recent study by the Idaho National Laboratory, comparing a NuScale plant with augmented steam compression and heating, to a high temperature gas reactor, has shown that both options are technical viable and economically competitive.

This paper examines the use of a steam production cycle in which steam generated by a single NuScale Power Module (NPM) is directed through an intermediate heat exchanger to produce process steam that is subsequently compressed and heated to achieve commercial scale steam temperatures, pressures, and flow rates. For example, a six module NuScale plant fully dedicated to steam production can produce 1088 metric tons of steam per hour (2.4 Mlb/hr) at 500^oC and 6.8 MPa using commercially available compressors and heaters. The NuScale flexible modular design makes it possible to assign one or more NPMs to produce steam and other NPMs to produce electricity, or each NPM can produce steam and electricity simultaneously. Process controls and regulatory requirements are also evaluated.

The Inertial Electrostatic Confinement Thruster (IECT) presents a promising solution as a space electric thruster device that employs an external centripetal electrostatic field to generate thrust through plasma interaction. This paper proposes a nuclear-powered IECT space propulsion system by using the system simulation language Modelica. To improve the accuracy and efficiency of system simulation across different spatiotemporal scales, this paper introduces a synchronous time Modelica-C coupled simulation method that accelerates calculations related to the IECT core. Furthermore, a multi-dimensional particle-in-cell simulation is implemented to better represent the physical processes occurring within the IECT core. System-level simulations are conducted to analyze the performance of the proposed system under various working conditions. The simulation results demonstrate that the proposed system can achieve satisfactory performance with significantly reduced resource requirements. Notably, Modelica exhibits robust capabilities for modeling space nuclear power systems and accurately describes plasma systems when coupled with external C code.

Integrated energy systems (IES) represent an emerging innovation for decarbonizing the power and industrial sectors. In response to this transition, decision-makers must address site-specific capacity and operation planning for heat and power supply, as well as the extent of heat and market engagement. The literature on the IES widely evaluates its economic viability under energy arbitrage operations. These operations leverage price differentials by reallocating energy production across spatial or temporal dimensions. However, prior studies have not examined various conflicting goals that decision-makers encounter in investment. For instance, industrial plant owners may aim to maximize nuclear heat utilization in their production processes to meet carbon emission targets, potentially replacing their existing fossil fuel energy sources. On the other hand, some may have limited grid access capacities for selling and buying electricity, which constrains arbitrage operations. Thus, we aim to make the following contributions in this work: (1) we introduce four reactor deployment scenarios to examine varying reactor capacity planning, considering decision-makers to either partially or completely replace their existing energy facilities with nuclear energy, (2) we formulate different grid access availabilities to identify optimal thermal energy storage (TES) and balance-of-plant capacities under site-specific constraints, (3) we assign additional operational targets, including maximizing electricity sales, minimizing carbon emissions, and minimizing dependency on external grids. The Xe-100 reactor and two-tank molten salt TES designs are optimized for various real-world industrial load scenarios. Our results reveal significant variations in system sizing and operation, highlighting the importance of including tailored constraints and operational goals.

Thermal energy storage is a promising technology that enables greater efficiency, load following, and therefore enhanced economy of nuclear reactors. The metrics to evaluate latent heat storage units are often material focused, which accounts for properties such as conductivity, latent heat, and density but fails to represent system level characteristics. We instead introduce a realistic metric that captures a more comprehensive system level thermal performance. To this end we seeked to use rate capability curves and a Ragone plot in the context of high-temperature latent heat storage. The Ragone plot helped elucidate design effectiveness and material pairings that would best balance energy storage and power delivery.

In this study, we designed a latent heat thermal storage system to pair Kairos Power’s KP-FHR with an energy storage amount equal to 10 hours of full power. Specifically, we investigated the impacts of thermal storage geometries and heat exchanger configurations on the system capacity and scale. Our preliminary design of the latent heat storage system has achieved a 7 times volume reduction compared to an equated two-tank sensible heat storage. We also carried out CFD simulation in Star-CCM+ to study the transient characteristics of these candidate geometries. Using these simulation results, we have demonstrated the use of rate capability curves and a Ragone plot to evaluate the comprehensive system performance.

The IAEA Coordinated Research Project (CRP) ‘Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop (NACIE)-CRP-I31038’, is a benchmark based on experimental data provided by the Heavy Liquid Metal (HLM) loop NACIE-UP (NAtural CIrculation Experiment- UPgraded) located at the ENEA Brasimone Research Centre. The primary system of the NACIE-UP facility consists in a rectangular loop which allows performing experimental campaigns in the field of thermal-hydraulics, fluid-dynamics and heat transfer of HLM. The primary loop is composed of two vertical pipes, working as riser and downcomer, hydraulically connected by two horizontal pipes. The facility includes also an ancillary gas system and a pressurized water secondary side for the heat removal from the primary loop. The test section for the experiments consists of a 19 electrically heated Fuel Pin Simulator (FPS) arranged in 3 ranks with a triangular pitch. The pins are placed on a hexagonal lattice by a suitable wrapper, while the wire spacer is adopted. The main objective of the performed experimental campaign was to perform integral system and local thermal-hydraulic analysis. Moreover, some of the performed tests were characterized by non-uniform heating of the bundle. The benchmark is divided in an open phase with cases ADP10 and ADP06 and a ‘blind’ phase with an active sector in the FPS ADP07. The benchmark is divided into 5 Work Package: WP1-System Thermal Hydraulics, WP2-Computational Fluid Dynamics, WP3-Subchannel Analysis, WP4-Multiscale Analysis, WP5- Uncertainty Quantification. In the paper, the different experimental test cases with boundary conditions are presented.

Benchmarking codes and methodologies against experimental data increases credibility of tools for liquid metal reactor design. Decades of experience have been gained during past and ongoing EU projects at KIT investigating liquid metal thermo-hydraulics. (CRP) ‘Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop (NACIE)-CRP-I31038’, is used for the verification of predictive capabilities of our modelling approach. The experiments test section consists of 19 electrically heated wire wrapped fuel pin simulator, arranged in 3 ranks with a triangular pitch. The benchmark offers in the open phase data for symmetric-heated forced and natural convection cases, which we analysed. These results showed very good local agreement and were published.

Appreciable effects of asymmetrical heating for forced and natural convection only become relevant in the new blind study presented here. The comparison of experimental data to all participants solutions will be presented in the NURETH21 in a separate paper. This work concentrates on the employed model, uncertainty quantification and comparison of our blind case results to the previous symmetric cases and experimental data.

The IAEA Coordinated Research Project (CRP) ‘Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop’, provides an opportunity to validate and improve thermal-hydraulic analysis codes used for simulating heavy liquid metal systems. The benchmark consists of two open cases to be used for verification of computational models – one each with uniform and non-uniform symmetric heating. After this first phase of the CRP, a blind case with asymmetric heating will be used to validate the accuracy of the models.

Within the CRP, one work package is devoted to Computational Fluid Dynamics (CFD) benchmark for the Fuel Pin Simulator (FPS), which represents a prototypical wire-spaced fuel pin bundle. Unlike system and subchannel codes, CFD is capable of resolving a detailed three-dimensional model of the FPS providing better a representation of the friction and heat transport phenomena involved, albeit at higher computational costs. The CFD benchmark is limited to the two steady states for each case, before and at the end of the forced-to-natural circulation transient. The benchmark participants use different CFD codes, implementing different sets of physical and/or numerical models. The present paper reports the collective CFD results obtained by the participants and comparisons with the experimental data. The comparison, here, is focused on temperature predictions for 67 thermocouple locations inside the FPS. The first phase of the benchmark provides an insight into the efficacy of different modelling strategies considered, and highlights the need of further investigations to improve the modelling of liquid metal-cooled wire-spaced bundles.

Recently there has been a notable increase in interest in Small Modular Reactors (SMRs) and Micro Modular Reactors (MMRs) due to their potential to improve energy supply reliability and reduce carbon emissions in isolated power grids. MMRs use high-temperature heat pipes, which typically employ liquid metals such as potassium or sodium as the working fluid, to extract heat from the reactor core. The MISHA research project, funded by BMBF, seeks to expand expertise in the application of heat pipes as the primary heat transfer mechanism in MMRs. This project includes the construction and the testing of full-scale high-temperature heat pipes using a newly established modular Heat Pipe Tester (HPT). The HPT's flexible design allows for testing heat pipes of different sizes and under various conditions, providing a thorough evaluation of their performance and advancing knowledge of heat pipe efficiency in various settings. Moreover, the experimental results will be used for the further development and validation of the GRS nuclear safety code system ATHLET.

The first heat pipe with reduced length of 2 m has been assembled, filled with potassium, and sealed. While the main HPT is still under construction, initial tests have been conducted using a smaller version of the tester. The heat pipe was tested at temperatures reaching up to 850°C, with up to 4 kW of power supplied. The behavior of the heat pipe during startup, steady-state operation, and cool-down phases was monitored and analyzed. Results have been compared to findings from other experiments documented in the literature.

This paper outlines recent and ongoing collaboration efforts between Canadian Nuclear Laboratories (CNL) and Korean Atomic Energy Research Institute (KAERI) on nuclear reactor safety research and development (R&D). The collaboration was motivated by a memorandum of understanding (MOU) between KAERI and Atomic Energy of Canada Limited (AECL) to work together in an innovative nuclear R&D partnership, with a broad focus including the safe deployment of small modular reactors (SMRs). The strategic drivers for SMR development and deployment in Canada are described by the Government of Canada’s SMR Roadmap and Action Plan (2018, 2020), which recognize that Canada needs viable and clean energy sources for different applications, and these needs could be supported by various types of SMRs. Korea’s SMR R&D priorities are described based on national programs actively promoting the securing of core technologies for the development of next-generation nuclear reactors. KAERI’s R&D covers a wide spectrum of scientific, engineering, and technical activities, supported by the utilization of large research facilities including the Advanced Thermal- hydraulic Test Loop for Accident Simulation (ATLAS), which is an integral effect test facility. The discussion herein focuses on topics of thermal-hydraulics for the development and safe deployment of SMRs, including appropriately scaled integral and separate effects experiments and analysis. Topics include core flow distribution, instabilities of two-phase natural circulation, and passive safety systems including the participation of CNL in the ATLAS-4 project led by KAERI starting in 2025, as well as the associated activities, schedules and expected outcomes.

Due to the advantages of efficiency and flexibility, floating nuclear power plants have become a focal point for research and development across various countries worldwide. In marine conditions, the movement of vessels alters the Critical Heat Flux (CHF) characteristics of nuclear reactors, which is essential to be reconsidered. In this paper, the CHF experiment, operated with R134a in the pressure range of 1.6-2.7 MPa and the mass flux range of 1000–3000 kg·m-2·s-1, has been conducted in both vertical and inclined conditions. The test section consists of a movable tube with an inner diameter of 8 mm and a heated length of 0.8 m or 1.6 m. The experimental results show that as the critical quality increases, the effect of inclination on CHF changes from reduction to no effect. Computational Fluid Dynamics (CFD) method was employed to simulate the experiment with inclination angles of 0° and 25°. The results indicate that the inclination causes a shift in the symmetrical distribution of the flow field, with a particularly significant impact observed on void fraction distribution. Bubbles tend to migrate towards the upper part of the inclined tube, leading to the accumulation of bubbles. Meanwhile, the liquid also supplements the upper wall. It may be the combined effect between the two that influences the reduction or invariability of the CHF in the inclined tube.

Accurate prediction of rod-bundle critical heat flux (CHF) remains a great challenge in evaluating the thermal safety margin of a nuclear reactor due to the lack of realistic models and experimental databases for the complex CHF phenomenon. This study examines the occurrence of CHF in a rod bundle using the MARS-KS subchannel analysis to verify CHF models and to provide useful supplements to CHF modeling. The examination uses the Wisconsin 2x2 rod bundle CHF experimental data. The CHF is detected by a rapid rise of the rod surface temperature with a step flow reduction or a step power increase, and the detected CHF values are compared with the measured values and with potential CHF mechanistic models (e.g., bubble crowding and sublayer dryout). The influence of radial/axial power distribution, space grid, cold wall, and bundle size on CHF is evaluated. Especially, CHF under the low flow low pressure is emphasized. The study is expected to provide a realistic methodology for evaluating CHF models based on more actual flow conditions and to broaden understanding of important factors affecting rod-bundle CHF.

Subchannel analysis plays an important role in nuclear reactor safety analysis, enabling better core thermal hydraulic predictions of parameters like the critical heat flux (CHF). This research focuses on developing a benchmark exercise for the COBRA-TF (CTF) computational code by performing subchannel analysis of the Electric Power Research Institute (EPRI) CHF database. The EPRI database comprises over eleven thousand experimental data points from rod bundles with diverse geometries, with uniform and non-uniform axial heat flux distributions. Operating conditions range widely, with pressures from 1 MPa to 17 MPa and mass fluxes from 50 kg/m²s to 6000 kg/m²s, providing a robust foundation for assessing CHF models. This benchmark aims to systematically compare the performance of widely used CHF correlations, including the Look-up table, Biasi, and W-3 correlations, under varying flow regimes and geometrical configurations. We aim to assess CTF flow regime and heat transfer models against CHF predictability. This work is expected to enhance the fidelity of CTF predictions and improve safety margin and performance evaluations in nuclear reactor design and operation. Furthermore, the benchmark will be a valuable resource for validating and refining CHF models.

The US nuclear industry is exploring an option to improve operational economics for the current fleet by seeking a cladding-performance based safety criteria as opposed to the current limit requiring complete avoidance of critical heat flux (CHF). Past experience has shown that not all events leading to a dryout are severe enough to cause fuel performance degradation. Allowing temporary dryout of the fuel, known as time-at-temperature (TaT), could allow for economic improvements to current plants without compromising fuel integrity. To support this effort, a comprehensive program is being executed by the US Department of Energy that includes generating cladding material data under TaT conditions, development of new mechanistic models, and demonstration of modeling and simulation capabilities for transients of interest. This paper presents current work done to assess the thermal-hydraulic subchannel code, CTF, which is a package used in the VERA core simulator, which will ultimately be used for TaT analysis. CTF will provide the T/H boundary conditions that will be needed for fuel performance analysis in the Bison code and it will therefore be necessary to quantify both the accuracy and uncertainty of post-CHF models. This paper presents the results of using the BFBT and Harwell tests for CTF validation, which both experience dryout conditions. This work also led to the implementation of an alternate post-CHF heat transfer modeling package that leads to improved agreement with experimental data. Agreement with experimental data for tube-geometry is generally good, but biases were detected for rod-bundle geometry that will require future improvements.

The CTF subchannel code which is used for the Thermal Hydraulic (T/H) solution in the Virtual Environment for Reactor Applications (VERA) is supporting the light water reactor application area in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Under this program, the ability of CTF is sought to be improved to model dry-out and re-wetting behavior in BWRs, which impacts its ability to model anticipated operational occurrence (AOO) transients using the time-at-temperature (TAT) approach which aims to demonstrate that the fuel rod’s integrity is not challenged during a mild elevated temperature transient. These models must be validated so that the thermal-hydraulic behavior can be used as boundary conditions in fuel performance codes. In order to improve CTF’s dryout location prediction and the post-dryout behavior prediction, the primary goal of this study is to perform sensitivity analysis based on SOBOL indices and other sensitivity analysis methods to identify the physical models which most affect the Figure of merit (FOM) in the flow regimes of interest. A multitude of tests will be used for model improvement such as the harwell tests, the BFBT turbine trip test, the FEBA tests etc., which are all part of the CTF V&V assessment suite, as well as expanding the test suite with the IFA613 tests. The second goal of the study is to perform model calibration using a Bayesian approach, which will also provide model uncertainty that will be used in a future uncertainty quantification study.

OpenSTREAM is a new open-source, one-dimensional, flexible computational environment designed to simulate boiling two-phase flows in single straight channels using various multi-field solvers ranging from the homogeneous equilibrium model to an advanced four-field model of annular two-phase flow. This paper applies OpenSTREAM’s three-field model to simulate a series of tests conducted at the Karlstein Thermal Hydraulic (KATHY) Test Loop under Boiling Water Reactor (BWR) conditions, including core instabilities. The 10´10 rod bundle geometry is represented in the code as a three-wall channel, accounting for (1) the adiabatic fuel shroud and central water channel, (2) the fuel rod with the highest radial power peaking factor, and (3) the remaining fuel rods. Initial simulations of single- and two-phase pressure drop tests are performed to calibrate the pressure loss coefficients of the spacer grids. A feature to account for enhanced droplet deposition downstream of the spacer grids is implemented in OpenSTREAM and calibrated against critical power tests. This feature enables accurate prediction of critical power and its associated elevation, determined by iterating the power until complete liquid film dryout is achieved anywhere on the hot rod. The simulation results show consistent agreement with the experimental data for the steady-state critical power across the range of tested boundary conditions. Preliminary transient simulations show that OpenSTREAM can predict dryout and rewet with time delays from inlet conditions representative of density waves.

With the advantages of high energy density, easy transportation and no pollution, hydrogen is a potential energy source for large-scale application to achieve near-zero carbon emissions, as well as an important industrial raw material. Since nuclear energy could provide stable and based loaded power with zero-carbon emissions, it is an ideal primary energy to produce hydrogen that is a kind of secondary energy. Currently, High Temperature Gas-cooled Reactor (HTGR) and water-cooled Small Module Reactor (SMR) could both be used for hydrogen production by steam reforming technology. In this work, a one-dimensional reaction flow model of the reformer tube, which is the core equipment in this methane-steam reforming hydrogen production using HTGR, was developed. The simulation results were compared to the latest experimental results, demonstrating the good validation. A comprehensive parametric sensitivity analysis on the reformer tube was performed using this model, providing a useful model to analyze and design a reformer tube for hydrogen production using HTGR. Additionally, a hydrogen production system with SMR coupled with methanol steam reforming was designed using SMR as the heat supply source. Parameter sensitivity analysis of this system was performed, and the optimal reaction conditions as well as the optimal reaction parameters were determined to provide guidance for nuclear hydrogen production system design. This work provides a general concept for nuclear hydrogen production by steam reforming technology.

An advanced concept of the Passive Molten Fast Salt Reactor (PMFR) has been recently proposed in the Republic of Korea as part of efforts to develop molten salt small modular reactors. Molten salt reactor (MSR) technologies have gained attention for their improved efficiency, enhanced safety, and capability for high-temperature operation, enabling non-electric process heat applications such as hydrogen production. A key innovation of the PMFR is the natural circulation of liquid fuel salt within the reactor loop, eliminating the need for pumps. This design improves safety by reducing reactor risks associated with pump reliability. Additionally, the PMFR is designed to integrate a compact and high-efficiency supercritical CO₂ (SCO₂) power conversion system. This study evaluates the feasibility and performance of hydrogen production systems coupled with the PMFR. For power generation, the study incorporates an SCO₂ Brayton cycle, recognized for its compact size and efficiency, and models its performance to optimize the use of the PMFR's thermal output. Potential hydrogen production methods analyzed include alkaline water electrolysis (AWE), polymer electrolyte membrane (PEM) electrolysis, and high-temperature steam electrolysis (HTSE). Thermodynamic models are developed for each production method to assess their integration with the PMFR's thermal and electrical outputs. Comparative analysis reveals that HTSE outperforms other methods in terms of efficiency and compatibility with the PMFR's high-temperature operation. The findings highlight the advantages of combining advanced nuclear reactor systems like the PMFR with HTSE for sustainable and efficient hydrogen production, offering valuable insights into future energy system designs.

This study presents an integrated solar and affordable, robust, compact (ARC) fusion reactor-driven integrated energy system for the production of electricity, freshwater, and hydrogen. The main aim of the study is to develop and evaluate non-electrical applications of the ARC fusion reactor integrated energy systems. Within the scope of this study, the integrated system consists of five subsystems, including an ARC fusion reactor, a concentrated solar power system, an open feedwater Rankine cycle, a multi-effect desalination system, and a cobalt-chlorine (Co-Cl) thermochemical cycle. The analyses of each subsystem and the overall system are assessed with the approaches of energy and exergy using the first and second laws of thermodynamics. The overall efficiencies of the integrated energy system are compared with the efficiencies of the original ARC fusion reactor design. Moreover, the Shomate heat capacity equation is employed while the calculations of the Co-Cl thermochemical cycle are carried out. The energy and exergy efficiencies of each subsystem are calculated. Consequently, the integrated energy system produces approximately 129.7 MW of electricity, 2040.7 tons/h of freshwater, and 1 mol of hydrogen per second, with 44.57% overall energy and 47.91% overall exergy efficiencies.

Space nuclear reactor systems utilizing heat pipes, which can effectively transfer heat without the need for pumps, have attracted significant attention as a viable solution for lunar applications. Korea Atomic Energy Research Institute (KAERI) has designed a sodium heat pipe incorporating a braided wick to enable flexibility in bending and using sodium as the working fluid. In this study, a test assembly consisting of six sodium heat pipes, each with a diameter of 1/2 inch and a length of 1 meter, was fabricated. The 25 cm evaporator section at the bottom, simulating a reactor core, was constructed using a graphite block and electric heaters within a helium chamber. The 50 cm adiabatic section was constructed using ceramic board insulation to encase each individual heat pipe, along with an insulation box to cover the adiabatic section of the entire assembly. To ensure precise cooling and heat transfer quantification for each heat pipe, the 25 cm condenser section was designed with a dual-cooling system comprising air and water cooling jackets. The thermal performance evaluation is conducted at temperatures exceeding 700°C, with the six heat pipes collectively transferring a total heat load of 3 kW. Experimental data obtained from this test will serve as a basis for validating the design codes of lunar nuclear reactor systems utilizing heat pipes.

In this study, thermal response and ablation tests of a containment system for nuclear batteries under re-entry aerothermal conditions were conducted using a 0.4-megawatt plasma jet test facility in Jeonbuk National University. Nuclear-powered thermoelectric generators have been utilized in space due to their ability to produce heat and electricity over extended periods through radioactive fuel decay, independent of solar flux. For the safe design of space nuclear reactors and radioisotope generators, the containment system must maintain its integrity around the radioactive heat sources even in the event of an accident. In the case of an atmospheric re-entry scenario, the containment system may fail due to exposure to the high-temperature atmosphere. Therefore, carbon-based thermal protection systems are attached to the containment system for nuclear-powered thermoelectric generators. According to a case study on re-entry conditions for nuclear batteries, the peak heat flux reaches 3.4 MW/m² with a recovery enthalpy of 11.4 MJ/kg. In this study, tests were conducted under conditions of a heat flux of 7.7 MW/m² and a recovery enthalpy of 13.9 MJ/kg. Test results showed that for a 20mm diameter carbon-carbon hemispherical sample, the ablation rate and surface temperature reached 0.04 mm/sec and 1800°C, respectively, over 120 seconds. This test data can serve as a critical database for developing an evaluation model for carbon-carbon thermal protection structures for nuclear-powered thermoelectric generators in space.

Thermionic reactors, with proven success in space applications and superior power scalability, present a promising technological solution for space nuclear power systems. To investigate the operational characteristics of thermionic space reactors, a system analysis code is developed based on the TOPAZ-II reactor. This code enables coupled nuclear-thermal-hydraulic-electrical calculations. The steady-state validation of the system analysis code is conducted according to the design values. To demonstrate the transient calculation capability of this code, the transient parameters during the start-up process are compared with the results of the referenced transient analysis model. The effects of cesium vapor pressure, nuclear power, and load resistance on system-level steady-state output characteristics are analyzed, and the intervals of the boundaries for optimizing system performance are determined. The results indicate that the steady-state calculation error of the developed code is less than 2.5%. The system responses during the start-up transient process agree well with the referenced values. The transient calculation of the system analysis code with comprehensive models is more consistent with the engineering practice. When the single-boundary variation intervals of cesium vapor pressure, nuclear power, and load resistance are (0.76 torr, 2 torr), (115 kW, 135 kW), and (0.068 Ω, 0.141 Ω) respectively, the system can achieve more efficient output electrical power than that in the benchmark steady-state. These findings provide valuable insights for improving the operational strategies of thermionic space reactors, and the system analysis code could serve as a theoretical tool for safety analysis.

Accurately modeling BWR core instability phenomena using coupled system codes presents significant challenges, particularly due to numerical diffusion in the calculations, which can dampen flow and power oscillations during the analyzed scenarios. This study examines the impact of numerical diffusion in the TRACEv5p9/PARCSv3.4.3 coupled code when modeling out-of-phase neutron flux oscillations. A 3D core model of a General Electric BWR/6 reactor is developed under flow-power conditions representative of a core stability test. To induce an instability scenario, an out-of-phase oscillation aligned with the first azimuthal mode is triggered through control rod maneuvers. The thermal-hydraulic core channel layout is defined based on the dominant Lambda modes of the reactor core, ensuring consistency between the modeled flow distribution and the expected oscillation patterns.

To mitigate numerical diffusion, TRACE incorporates several second-order numerical schemes. A comparative analysis is conducted between the default first-order upwind scheme and higher-order methods to evaluate their impact on numerical accuracy. The results highlight the importance of selecting an appropriate numerical scheme to minimize diffusion effects and improve the predictive capabilities of instability behavior in BWR reactors.

There has been a growing demand for floating nuclear power plants (FNPPs) to reduce greenhouse gas emissions and provide remote energy supply in recent years. Unlike conventional land-based nuclear power plants, FNPPs experience continuous changes in heat transfer and flow characteristics due to ocean motion. In this context, the effect of ocean motion on the critical heat flux (CHF) has been studied. However, the available studies are limited, and the range often differs from the operational range of nuclear power plants. In addition, the nuclear fuel used in nuclear power plants has a cosine-shaped axial power distribution; however, experimental studies under ocean motion reflecting a cosine-shaped axial power distribution are needed.

This study conducted a flow boiling CHF experiment using the NEOUL-H platform, capable of simulating heaving motion. To simulate the ocean environment, the experiment was conducted with a period of 3-6 seconds and a maximum acceleration of 0.6 g. The CHF test loop used R134a as the working fluid, and the experiment was conducted under thermal-hydraulic conditions corresponding to PWR operating conditions through fluid-to-fluid scaling. The test section consists of a single heater rod with annular channel, and the axial power profile of the heater is cosine-shaped.

In the experiments, CHF was measured under static and heaving conditions. In the heaving condition, CHF decreased compared to the corresponding static condition. We also found that the magnitude of CHF variation depended on the thermal-hydraulic conditions, such as mass flux and pressure, and the heaving conditions, such as period and amplitude.

Three-period minimal surface (TPMS) heat exchangers have great potential in nuclear engineering because of their compact design and excellent thermal and physical properties. To further improve the performance of heat transfer capability, control factors α and β are introduced based on the standard Gyroid function to regulate the closure of the "through-hole" structure and the surface "fold" microstructure, and the flow heat transfer characteristics of the modified Gyroid TPMS structure heat exchanger are investigated based on numerical simulation and experimental measurements. The results show that with the increase of α value, the extreme temperature (Tmax) of the Gyroid structure decreases by 16.1-27.9 K, the convective heat transfer coefficient (h) increases by 28.3-33%, and the Nussel number (Nu) increases by 1.4-3.2%. With the increase of β value, the extreme temperature (Tmax) of the Gyroid structure decreased by 4.9-7.4 K, the convective heat transfer coefficient (h) increased by 4.3-8.2%, and the Nussel number (Nu) increased by 0.7-3.5%. The "through hole" closure and the addition of the surface "fold" microstructure significantly improve the convective heat transfer performance of the TPMS structure and increase the pressure drop. Taking the comprehensive performance evaluation factor (PEC) as the evaluation index, it is recommended to choose α=2.0 or β=0.8 to achieve the best effect. The scheme and structure of this study can provide a new idea for the further improvement of the TPMS structure heat exchanger.

Fluid-structure interaction (FSI) is commonly seen in nuclear power plants, such as fluid flow in piping systems and steam generator tube. While FSI analysis methods based on finite element methods (FEM) are widely used, they face difficulties in handling structural fractures and large deformations due to the complexity of mesh re-generation. Lagrangian particle methods offer a promising alternative, enabling stable computation of these phenomena. However, challenges remain, such as conservation in the discretization system which is important for stable simulation.

To address these challenges, this study presents a novel Lagrangian-Lagrangian FSI solver with a physically consistent particle method moving particle hydrodynamics (MPH-MPH). The governing equations for elastic bodies and incompressible flows are discretized using the Moving Particle Hydrodynamics (MPH) method.

Several benchmark tests, including single bar vibration, a hydrostatic water column on an elastic plate, and a dam break with an elastic gate, verified and validated the method’s accuracy. The calculation results had good agreement with theoretical predictions and experimental data.

In summary, the MPH-MPH method shows significant potential for solving FSI problems involving large deformations and fractures.

The enhancement of Latent Heat Thermal Energy Storage (LHTES) systems through fin geometry optimization remains a critical challenge for leveraging the full potential of renewable energy sources. This study focuses on optimizing the geometries of tree-shaped fins to enhance power and energy densities in LHTES systems. The goal is to find branch designs with high energy and power density through a novel surrogate model-based optimization strategy that explores a broad design space. The surrogate models applied, including linear regression, principal component analysis-based linear regression, artificial neural networks, and random forest, are evaluated for their predictive performance. The random forest model demonstrates superior accuracy in predicting targets. The optimization process results in a Pareto-optimal design with a volume fraction of 33.9%. This optimal design substantially enhances the system's power density by 61.6% compared to conventional plate fins at an equivalent energy density. This optimized design improves energy and power density, achieving a uniform end-to-branch distribution, which is a pivotal factor for consistent temperature distribution and improved thermal efficiency. By integrating surrogate-based optimization with broad ranges of the tree-shaped fin design, this research has significantly improved the operational efficiency of LHTES systems. This research promises more effective thermal management and provides a methodological framework for design innovation in thermal energy storage.

The pursuit of sustainable and clean energy has intensified global interest in nuclear fusion, particularly through tokamak reactors, which are seen as promising candidates for safe and efficient energy generation. A crucial aspect of their development is the study of the Balance of Plant (BoP). It includes all the systems belonging to the main heat transfer chain, devoted to the plasma power removal and to its conversion into electricity. This research involves BoP thermal-hydraulic analyses, focusing on the cooling system behavior, carried out by using RELAP5 system thermal-hydraulic code.

Given the pulsed nature of tokamak operation, alternating plasma pulses and dwell phases, special attention is given to transient analysis. These transient conditions pose challenges to reactor operations, influencing its efficiency, thermal stability, and safety. For this, the study includes an investigation of thermal storage systems designed to accommodate plasma power variations. A study of Intermediate Heat Transfer System (IHTS) design is proposed, which would couple the primary circuit with the Power Conversion System (PCS).

The primary goals are to evaluate the BoP response under both normal and off-normal conditions, including potential accident scenarios, to assess the plant operational efficiency and safety margins. This not only improves the understanding of energy transfer and heat management in fusion reactors but also offers information for the design and implementation of safety measures. This research contributes to optimizing tokamak design and operation, while providing pre-conceptual designs for main cooling system components. These insights and data aim to support the development of tokamak fusion reactors.