Conference Agenda

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

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

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

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

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

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

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

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

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

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

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

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

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

Nuclear energy provides about 70% of France’s electricity, with 56 pressurized water reactors (PWRs) operated by Electricité de France (EDF). EDF R&D uses advanced fluid mechanics to ensure reactor safety, employing in-house 3D codes like neptune_cfd to study two-phase flows and critical phenomena such as the boiling crisis.

This article focuses on phase change at the interface between liquid and vapour, often referred to as bulk or interfacial condensation/boiling. Although overshadowed by wall-driven condensation/boiling, interfacial phase change is crucial in nuclear applications, particularly in liquid metal flows for sodium-cooled reactors and microchannels used in the nuclear industry.

While single-fluid Volume of Fluid (VOF) codes effectively model interfacial phase change, multi-field computational fluid dynamics (CFD) codes lag behind. This article introduces a new phase change model for multi-field codes, using the gradient method to capture interfacial phase change accurately. The model is validated against both analytical and experimental cases involving bulk boiling, showing excellent agreement. Its mesh convergence aligns with single-fluid codes, and we propose a hybrid approach combining this model’s accuracy with the computational efficiency of dispersed-phase methods for simulating complex two-phase flows.

A novel analytical solution-based thermal resistance network computational method has been proposed to provide a more accurate and reasonable temperature calculation framework for pebble bed reactors. This method generates positional and contact information for large-scale pebble beds based on the analytical solution of multidimensional generalized thermal resistance and results from the discrete element method. It also calculates the generalized thermal resistance between the center points of adjacent particle contact surfaces. The generated data is trained using decision tree and random forest algorithms, constructing multiple weak classifiers (i.e., individual decision trees) and combining them into a strong classifier to reduce overfitting and enhance the model's generalization capability. A random forest model was built on the TreeBagger framework, utilizing 100 decision trees and 3 leaf nodes. The importance of each feature's impact on thermal resistance was analyzed, and the trained values effectively reflected the thermal resistance values, achieving a maximum percentage error of 1.45% in the testing set. Validation was conducted using simple cubic, body-centered cubic, and face-centered cubic packing arrangements, showing good agreement with finite volume method results. The novel thermal resistance network model was applied to compute the temperature field caused by heat conduction in the HTTU, confirming the model's feasibility and providing the temperature distribution at various nodes.

The accumulation of Corrosion-Related Unidentified Deposit (CRUD) in the core of the Pressurized Water Reactor (PWR) poses potential threats to reactor safety. This study investigates the deposition behavior of CRUD on the PWR fuel cladding surface, constructing a high-precision 3D CRUD dynamic growth model within a complex coupling framework. The Discrete Phase Model (DPM) is employed to analyze the transport and deposition processes of particulate corrosion products within the fuel rod bundle channels. By integrating the Zeta potential and Electrical Double Layer (EDL) model, the study systematically examines how the Zeta potential near the fuel cladding influences particle deposition behavior. Dynamic mesh technology is used to visualize the dynamic growth process of CRUD layer. In parallel, species transport equations are employed to analyze the distribution characteristics of metal ion concentrations and pH within the coolant. Finally, a multi-physics coupling mechanism of CRUD deposition behavior with rods channel inclusion, water chemistry, boiling heat transfer, and flow field distribution is revealed. The results show that Zeta potential and local pH affect the deposition behavior of corrosion product particles. Specifically, when the Zeta potential of the fuel cladding wall is positive, deposition becomes more challenging in acidic conditions but easier in alkaline environments as the Zeta potential increases. Furthermore, the rough surface of the CRUD layer induces localized accelerated flow near the cladding wall, which exacerbates electrochemical corrosion. The complex coupling effects of CRUD layer thickness and temperature field, particle deposition behavior with Zeta potential, local accelerated flow and electrochemical corrosion are revealed.

Flows past a solid wall for a well-known region between the wall and the bulk of the flow, the boundary region. The characteristic thickness of this boundary region is defined by the appropriate diffusion coefficients and is a place of high gradients and non-linear behaviour. In simulations, we prefer to avoid the calculation of the flow properties in the inner boundary region, since it increases the computational cost of the simulation greatly, being a very thin region next to the walls. This is possible, as the inner boundary layer exhibits a self-similar behaviour, that can be described with explicit functions, called wall- functions.

For thermal fields, the temperature gradient at the wall-normal direction determines the heat extracted from the wall, therefore its correct representation will determine the overall temperature field in the domain. It is therefore important we accurately compensate for the effect of the wall on the rest of the flow, if not resolved.

In the proposed paper we examine a turbulent boundary layer with a DNS with multiple temperature fields of various Prandtl numbers to design more accurate thermal wall-functions. The simulations are performed by the incompressible Navier-Stokes solver Nek5000 and restricted to forced convection flows. We will use these simulations to establish highly accurate explicit wall function that depends on the Reynolds and Prandtl number, making it applicable for a wider range of fluid flows.

This paper presents the 3D modelling and CFD analysis of a CANDU irradiated fuel bay (IFB). CANDU IFBs are significantly different from light water reactor spent fuel pools in terms of the bundle type and orientation of the assemblies. Following the Fukushima Daiichi accident, the Canadian Nuclear Safety Commission (Canadian regulator) undertook a vigorous re-evaluation of the current safety measures and margins of the CANDU IFBs under various stages of an extreme beyond design basis accident scenario. Amongst a few phenomena and hypothetical scenarios of interest, advancing the knowledge of the air-cooling effect on the fuel assemblies and storage racks during the complete loss-of-coolant accident (LOCA) was deemed significant. Under the pan-Canadian PIRT effort, the air-cooling effect on the fuel assemblies, which is driven by the natural convection and radiation modes of heat transfer, was identified as an area of high-importance with low knowledge-level phenomenon for IFBs. The objective of this study is to simulate the airflow and temperature distribution around the irradiated fuel racks at various decay power under a postulated accident scenario (complete LOCA) using the CFD code Simcenter STAR‑CCM+. A detailed 3D model based on the geometry of the CANDU fuel storage module was developed that was qualitatively analyzed in CFD for its capabilities to predict the sheath temperature of the irradiated fuel; a key parameter to monitor the severity of the IFB LOCA. It is anticipated that the developed CFD model could be leveraged to inform lower-fidelity codes and to guide experiments to develop validation data.

In molten salt reactor (MSR) systems, tritium is generated in significantly larger quantities when compared to other reactors. Due to tritium’s high permeability and solubility, tritium must be removed from the molten salts to reduce the associated radiological risk. Gas sparging is used for the online separation and removal of fission products within FLiBe where noble gas is bubbled through the molten salt to remove dissolvable fission gasses. To advance gas sparging technology, we are doing experiments to develop a detailed understanding of gas sparging behavior and create benchmark datasets for Multiphysics Object Oriented Simulation Environment (MOOSE) models. To do this, we have designed and built the Modular Separate Effects Test Facility (MSEFT) - Tritium Removal Investigation of Transport Interactions Using Mass-transfer (TRITIUM) flow loop to observe and measure localized bubbling behavior and integral gas concentrations using surrogate fluids. The Normalized Dissolved Oxygen Concentration (NDOC) within the surrogate fluid will be reported on for different prototypical conditions. The NDOC is used to characterize the liquid-gas mass transfer coefficients of sparging bubbles within water at various glycerol weight percentages used to match relevant non-dimensional numbers of Reynolds, Sherwood, and Weber. The NDOC data will be combined with measurements of relevant local bubble dynamics including average bubble diameter and velocimetry taken with high-speed shadowgraphy and Particle Image Velocity systems. These combined measurements will be useful to inform future design improvements for different MSRs’ gas sparging components.

In prior studies, we introduced an optimized prototypical natural circulation water-based reactor cavity cooling system (RCCS) design for a pebble-bed generic FHR based on 1D modeling and in later work performed more detailed 2D transient performance analyses. The prototypical design was an initial step to facilitate experimentation. Experiments in a university laboratory necessitate downscaling of the prototypical RCCS while maintaining key non-dimensional parameters such as Grashof number, particularly in systems involving natural circulation. The design of the RCCS, core, and the radiated power significantly affect the non-dimensional parameters in the RCCS. Tradeoffs exist between the non-dimensional parameters as a combination of design parameters may yield ideal scaling for one parameter but result in unacceptable scaling for other parameters. In this work, we systematically study the dependence of the relative scaling of the non-dimensional parameters compared to the prototypical case as a function of design and physics parameters using an iteratively refined base case design. These non-dimensional parameters considered herein include Grashof, Reynolds, Nusselt, Prandtl, Rayleigh, Biot, Stanton, Froude, and Richardson numbers. An idealized, downscaled design was obtained based on these analyses and a practical experimental setup was subsequently designed.

Kairos Power recently obtained NRC approval for construction of its high-temperature fluoride-salt cooled pebble bed reactor, Hermes, and is presently working towards submission of its operating license. As part of this activity, Kairos is using the KP-SAM systems code to perform safety analysis of the reactor under transient conditions. As part of the verification & validation (V&V) of this code, Kairos has built and operated a reduced-scale integral effects test (IET) facility which leverages hierarchical two-tiered scaling (H2TS) to simulate a loss of forced circulation (LOFC) in the reactor in which a pump failure leads to the onset of natural circulation. This facility employs a surrogate heat transfer fluid which simultaneously matches several dimensionless numbers – such as the Prandtl number – of molten salt at a significantly reduced temperature and scale, enabling rapid testing with high-fidelity instrumentation. This paper discusses the scaling, testing campaign, results, and future plans for the IET.

We present selected integral effect test results conducted in the STELLA 2 facility on the decay heat removal systems (DHRS), along with numerical analyses with MARS LMR system thermal hydraulics code. STELLA 2 is a large scale sodium test platform modeled after the Korean sodium cooled fast reactor, PGSFR, with a 1/5 scale reduction in length. The facility aims to evaluate the overall plant dynamics and safety aspects of the reactor under long term transient conditions. With equivalent conservation of the reactor components’ shapes and layouts, reactor transients can be investigated while maintaining the integral behavior and interactions of the prototypic reactor’s heat transport systems with minimal distortions. The facility has four individual DHRS loops, comprising two active and two passive loops differentiated by their cooling mechanisms for ultimate heat sinks. The selected tests focus on scenarios where the primary system after reactor shutdown relies solely on two DHRS loops, postulating a loss of offsite power condition. Overall, the system code successfully captured the general transient behaviors in the individual loops, although variations were observed in the peak temperature and the time to reach it, compared to the experimental data. It is believed that discrepancies in local temperature distributions, particularly in larger volumes, are attributed to the system code’s limitations in discretization method and relations.

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

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

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

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

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

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

The IAEA has been a key supporter of the development of advanced power-reactor technologies and nuclear fuel technologies for many decades. Its efforts include providing platforms for information exchange, organizing meetings, issuing publications, coordinating research activities and maintaining databases (for advanced reactor designs, fuels, fuel cycle and post irradiation examination facilities).

This presentation will highlight the IAEA’s on-going programmes and near-term plans to support the development of new reactor technologies and advanced fuels for both operating and innovative power reactors. This includes IAEA’s efforts in developing accident tolerant and advanced technology fuels (ATF), fuels for recycling/multi-recycling of nuclear materials, and advanced fuels for GEN-IV and small modular reactors, as well as advanced reactor designs.

Special emphasis will be paid to key Coordinated Research Projects (CRPs) including “Testing and Simulation of Advanced Technology and Accident Tolerant Fuels (ATF-TS)”, “Fuel Materials for Fast Reactors”, “Standardization of Subsized Specimens for PIE and Advanced Characterization for SMR and Advanced Applications”, “Fuel Modelling Exercises for Coated Particle Fuel for Advanced Reactors Including Small Modular Reactors”, “Developing a Phenomena Identification and Ranking Table and a Validation Matrix, and Performing a Benchmark for In-Vessel Melt Retention”, and “Advancing Thermal-Hydraulic Models and Predictive Tools for Design of SCWR Prototypes”.

IAEA Member States are strongly encouraged to participate in the IAEA’s topical meetings and new CRPs, which provide unique opportunities to engage with cutting-edge reactor and fuel technologies critical to the future of nuclear energy.

CRUD is one of the major considerations from the perspective of operation and safety, especially in PWR nuclear power plants. CRUD, which consists of corrosion products in the reactor coolant system, is known to induce thermal resistance, distortion of power distribution, local corrosion, boron hide-out, etc. Some of these adverse effects imply that the CRUD can affect the plant economics and core integrity hindering the thermal limit of nuclear power plants. These kinds of challenges significantly highlight that the CRUD effect should be investigated to calculate accurate operational margin of nuclear power plants in operation as well as in development. Various academic efforts have been made to figure out the mechanical and chemistry characteristics related to the deposition mechanisms and implications, trying to reflect them on the nuclear power plants. A reliable database of CRUD is necessary for solutions since deposit experiments are too hard to generate quantitative results under high-pressure/high-temperature PWR operational conditions. Hence, this paper investigates the CRUD effects utilizing the thermal properties obtained from experiments under actual PWR conditions. In this method, the CRUD layer is arranged on the surface of the fuel clad composing an active core in a simulated PWR plant model. Based on the results, the guideline can be made to calculate the local heat flux on the nuclear fuel considering the high burn-up rate of the reactor core. Further research will be conducted for the better quality of the database, expanding the test conditions and results.

Passive safety systems are widely used in the advance nuclear reactors. As an important component in the passive safety system, the core makeup tank (CMT) plays a key role in the safety injection and the core decay heat removal during the transient process of small-break loss of coolant accident (SBLOCA) . In this study, the thermal hydraulic behaviors of the CMT injection process were investigated by simulating different accident scenarios of the ACME test facility with different break sizes which includes the 1inch，2inch，4inch and 8inch breaks. Through the comparative analysis of the transient simulation under different break conditions along with visualization results, the switching process between the two working modes of CMT and its interaction with other safety injection component (such as ACC) were studied. Moreover, the mechanism of thermal stratification in CMT and the related influencing factors of CMT safety injection were analyzed. This work can provide guidance for the safety design and performance qualification of advanced passive nuclear reactors.

Abstract – In the framework of the OECD/NEA experimental projects, like ATLAS-3, a set of DEC-A experiments were carried out aiming at assessing the nuclear power plants capabilities to deal with complex accidental scenarios and evaluating the design provisions of the safety systems and the adequacy of the accident management measures. DEC-A scenarios are generally based on events and combinations of events which may lead to severe fuel damage in the core. The safety assessments are normally carried out using best estimate tools with the objective to demonstrate the fulfilment of the safety criteria and the design robustness. In this paper, a DEC-A experimental scenario carried out in the ATLAS test facility is considered. The ATLAS C3.2 test concerns a steam line break scenario relying on the passive auxiliary feedwater and operator action to cooldown the primary system. The transient involves complex interacting natural circulation phenomena including natural circulation flow interruption, steam condensation and heat exchange in large pool. In this framework the CATHARE code is used to simulate the course of the transient and the related natural circulation phenomena. It is shown, on the one hand, that the safety features of the design together with the operator actions are capable to bring the primary system to a safe end state and on the other hand, the CATHARE code prediction capabilities, for such complex scenario, are generally good. Nevertheless, additional efforts should be carried out to enhance the simulation under passive natural circulation cooling conditions.

During a Loss of Coolant Accident (LOCA) with failure of Emergency Coolant Injection (ECI) in a Pressurized Heavy Water Reactor (PHWR), the convective cooling is compromised, leading to an increase in the fuel channel temperature. Initially, the fuel temperature rises due to the decay heat and the energy stored in the fuel. The chemical reaction between the cladding and steam further escalates the temperature, causing the cladding to embrittle from oxygen diffusion. This can result in cladding rupture and the release of fissile materials. Additionally, this reaction produces hydrogen gas, which threatens the structural integrity of the containment. The heat from the fuel bundle is transferred to the Pressure Tube (PT) and the rising temperature of the PT leads to deformation, such as ballooning, sagging, or both, due to the rapid degradation of its thermo-mechanical properties, influenced by internal pressure. Given the significant risks associated with such accidents, it is crucial to study the behavior of fuel channels under LOCA conditions. This paper investigates the thermal performance of Indian PHWR under an oxidizing environment that simulates a late-phase accident scenario. The temperature profiles of the fuel element simulators, PT, and CT under steady-state conditions are obtained. A 37-element fuel bundle simulator is used, with the PT mounted eccentrically inside the CT, maintaining a 4 mm eccentricity.

Helical cruciform fuel (HCF), a novel nuclear fuel design, shows potential for enhancing power output and extending the service life of light water reactors (LWRs). While thermal-hydraulic studies on HCF assemblies are common, coupled analyses with high-fidelity neutron physics remain limited. This study establishes a CFD model of a 3×3 HCF assembly, integrating volumetric heat sources derived from neutron physics calculations to investigate flow and heat transfer phenomena.Key findings include helical variations in heat flux density (q) and wall temperature (T_w) along the flow direction. Due to the gap effect, q is lower in valley regions compared to blade regions, while T_w shows the opposite trend. Bulk temperature (T_l), however, lacks noticeable helical patterns. Under different heat source conditions with identical total power, peak values and positions of q, T_w, and T_l vary significantly. Condition one results in 70% higher q, an 18.8 K rise in T_w, and an 8 K increase in T_l compared to condition Two, with peaks occurring in different axial regions. Conversely, condition Two shows minimal axial q variation, with T_w and T_l peaking at the outlet. These results suggest a higher likelihood of boiling crises under condition one. Increased local axial power exacerbates circumferential non-uniformity in q and T_w, with heat transfer deteriorating at blade regions aligned with Twist angles of multiples of 90°, marked by reduced q and elevated T_w and T_l.

Fuel elements play a vital role in the safety and economic efficiency of nuclear power plants as one of the core components of a reactor. In addition to facilitating inter-channnel mixing between rods and enhances heat transfer, the wire-wrapped rod bundle is free from the supporting of spacer grid due to its self-locating structure through the contact between adjacent rods, which making it popular in recent research on reactor structural design. However, complex mechanical interactions often occur in wire-wrapped fuel rods under the constraints of reactor irradiation and high temperature, leading to stress concentration at the contact points of adjacent rods. This can easily cause fatigue damage to the fuel clad and affect its integrity. This study establishes a finite element model of wire-wrapped fuel rods by using the Ansys Workbench, taking into account the effects of irradiation, high temperature, and the geometric structure of the wire. The mechanical interaction characteristics of wire-wrapped rods under complex working conditions are investigated. The results obtained from this study on the mechanical characteristics of wire-wrapped rods can provide insights for the structural optimization design of fuel rods.

Externally cooled dead-end pipes, thermally and hydraulically connected to a hot source, exhibit complex physics, primarily driven by the interaction between penetrating swirl at the open end and stratified flow near the closed end of the pipe. These inherent instabilities in practice can lead to temperature fluctuations, potentially causing thermal fatigue and leakage in stainless-steel pipes. Such uninsulated pipes may be found, for example, in some 2 loop Westinghouse PVRs, where the safety injection (SI) pipes are connected directly to Reactor pressure Vessel (RPV). To better understand the thermal-hydraulic behaviour of this SI pipe configuration, CFD simulations were conducted. Despite advancements in computational power, such industry-level simulations remain challenging due to numerous uncertainties affecting prediction accuracy. Our study highlights that prediction of the turbulent swirl plus competing with the natural circulation is highly sensitive to CFD model settings, including boundary conditions, mesh resolution, turbulence models, and numerical methods (e.g., discretization schemes, solver types). These sensitivities suggest that the phenomena are unstable and chaotic. External air cooling, which induces flow stratification, emerges as the primary source of uncertainty, while unknown geometry at the inner edge of the DVI nozzle where the swirl forms, adds further complexity. Additionally, the large geometrical model restricts a systematic mesh sensitivity study, and the lack of sufficient experimental data limits the validation of turbulence modelling. All the above-mentioned aspects will be systematically reviewed in our paper, and supported by the CFD examples.

Integral effect test data obtained from the High Temperature Test Facility (HTTF) are being used for benchmarking CFD and system codes in the OECD-NEA Thermal Hydraulics Code Validation Benchmark for High-Temperature Gas-Cooled Reactors using HTTF Data. Five system codes SAM, RELAP5-3D, GAMMA+, SPECTRA, ARIANT, and CATHARE are used to model benchmark Problem 3 Exercises 1C and 1D, which simulate the steady state and pressurized conduction cooldown (PCC) transient of HTTF Test PG-27. This test examines the PCC phenomena progression in an integral test facility scaled to the General Atomics MHTGR design. The proposed exercises include well defined boundary conditions and assumptions so that code-to-code comparisons will help identify differences between modeling approaches, numerical methods, and uncertainties in the solutions of different codes. Preliminary assessment of the results shows that in steady-state operating condition, the codes agree very well for key parameters such as coolant temperature, solid temperature and flow distribution in core regions. In PCC transient, the agreement is reasonably good. The codes predict that natural circulation is several orders of magnitude lower than steady-state flow rate and core-wise heat transfer is therefore dominated by thermal conduction and radiation. The codes also predict similar temperature trends in the solid structures but there are discrepancies in the transient behavior. This is not surprising considering the vastly different modeling schemes of a very complex core geometry.

Extreme accidental scenarios in nuclear power plants (NPPs) may involve hydrogen formation and its pressurized release into the containment building, potentially leading to unintended explosions. A fundamental understanding of the complex physical mechanisms associated with such scenarios is critical for their prevention and mitigation. This includes investigating hydrogen jet dynamics during the initial stages of release under high-pressure conditions, which are relevant for hydrogen storage systems in nuclear facilities.

This work uses high-fidelity 3-D numerical simulations based on the Implicit Large Eddy Simulation (ILES) technique to investigate the turbulent characteristics and mixing of underexpanded jets, varying initial pressure ratios, and jet diameters. First, nitrogen jets released into atmospheric nitrogen are investigated, examining pressure ratios of 60, 30, 15, and 7.5 for a 3 mm diameter jet. This serves as a simpler case to analyze the jet flow dynamics. Second, we focus on high-pressure hydrogen jets released into air with the same pressure ratios and different jet diameters of 1.5, 3, and 6 mm to represent a reactor-scale problem. Both cases are validated against experimental data with an excellent agreement.

Key findings include the influence of pressure ratio and jet diameter on the turbulent jet self-similarity and mixing shear layer dynamics. A larger jet diameter enhances self-similarity, while a decrease in pressure ratio disrupts it. Higher pressure ratios result in thicker shear layers and broader temperature ranges. These insights contribute to enhancing safety procedures and protocols in nuclear systems and other high-pressure hydrogen storage applications.

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

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

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

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

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

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

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

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

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

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

The analysis of severe accidents in nuclear power plants is critical due to their potentially catastrophic impacts on public safety and the environment, underscoring the need for severe accident analysis codes like MELCOR. However, MELCOR faces two major challenges: (1) difficulty in solving multi-physics problems and (2) instability caused by complex computational schemes. To address these issues, this study investigates the feasibility of a physics informed neural network (PINN)-based MELCOR code by designing and evaluating a module for the CVH/FL package. MELCOR's governing equations were first implemented as a Python-based module for a thorough understanding, followed by the development of a PINN-based module applied to simplified 2-tank and 3-tank gravity problems. While both models approximated height and velocity well across most regions, discrepancies emerged when the height of the last tank approached the pipe. Notably, the PINN module struggles to accurately predict physical phenomena, particularly in scenarios involving singularities. We believe that our benchmarking study of PINN modules against the MELCOR CVH/FL package is very useful for examining its feasibility in severe accident analysis.

Accurate prediction of thermodynamic properties near the critical point is crucial for safety analysis in nuclear reactors, especially during severe accidents involving steam explosions. Existing methods face challenges in this region due to rapid and nonlinear changes in physical properties, leading to numerical instability and unreliable results. To address these limitations, we developed a deep learning-based standalone model that predicts physical properties near the critical point with high accuracy and computational efficiency. Utilizing data from the International Association for the Properties of Water and Steam (IAPWS), the model is trained to take specific internal energy and specific volume as inputs and outputs the corresponding pressure and temperature. The neural network employs a multilayer perceptron architecture with Leaky ReLU activation functions and is optimized using the mean squared error loss function and the Adam optimizer. Hyperparameter tuning, including adjustments to batch size and learning rate, was performed to enhance model performance. The developed model successfully captures the complex thermodynamic behavior near the critical point, overcoming the deficiencies of previous approaches. Integration of this model into the SCAR (Steam Explosion Code for Associated Risk) module, which is currently under development, enhances its predictive capabilities, providing more reliable inputs for severe accident analysis. This work demonstrates the potential of deep learning approaches in improving thermodynamic property predictions and paves the way for their application in other areas of nuclear safety analysis.

This study aims to reinforce safety measures for pressurized water reactors, ensuring more effective mitigation of severe accidents. Through uncertainty and sensitivity analyses, the research evaluates the effectiveness of reactor coolant system and steam generator injection strategies during an Extended Station Blackout scenario in the OPR-1000 nuclear reactor. Uncertainty analysis focuses on both code-related uncertainty parameters and the human reliability of executing time, critical factors that influence accident mitigation. Additionally, sensitivity analysis is performed to examine the injection rate of the Multi-barrier Accident Coping Strategies equipment, providing insights into the optimization of mobile equipment performance. The research evaluates the effects of these variables on key outcomes, including core cooling, reactor coolant system depressurization, and integrity of the reactor vessel. Utilizing the MAAP5 code, the study provides relevant data to enhance Severe Accident Management Guidance and improve accident management strategies.

As an integrated SMR, IP200 has the advantages of compact structure and high safety, and can be applied to floating nuclear power platforms through certain improved designs. The inherent characteristics and safety facility design of IP200 make its accident sequence slightly different from that of land-based PWR. The complete accident process from the initiation of the accident to the early occurrence of the reactor phenomenon, and then to the IVR and even the reactor reaction after the pressure vessel damage in the late stage of the serious accident, as well as the thermal and hydraulic effects of the safety facility input, are worth further research. A complete and detailed simulation model including the main coolant system and safety facilities of severe accident is established based on the mechanical severe accident analysis program Melcor and the integrated PWR thermal model, and the floating nuclear power platform IP200 is taken as the research object. The research results indicate the complete accident development sequence, key physical response characteristics of the core, and response characteristics of thermal and hydraulic parameters inside and outside the floating nuclear power platform under the condition of DVI pipeline rupture accidents before and after the failure of safety facilities, verifying the effectiveness of safety facility design.

One of the challenges in severe accident evaluation of Generation IV Sodium-cooled Fast Reactors (SFR) is the eutectic reaction between boron carbide (B4C) and stainless steel (SS), leading to boron migration in a molten pool within the core, which increases neutron absorption. To investigate this phenomenon, high-resolution radiative heating was employed to observe boron migration, eutectic behaviour, and melt structure. Experiments replicating control rod designs were conducted using B4C pellets in SS tubes at temperatures up to 1372°C. Two melting mechanisms were identified: SS separating from the B4C pellet and forming a melt drop, and B4C pellets fracturing due to thermal stress The use of visualisation techniques allowed for the detection of the eutectic onset, and the resulting eutectic melt was further analysed using material characterisation techniques. X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS) confirmed the formation of metal borides and metal carbides, attributed to the high chromium, iron, and carbon content. The current paper's findings confirm the relocation of the B4C-SS eutectic mixture and the formation of diverse boride phases, conditions likely to occur under extreme reactor conditions.

The cold trap of a liquid sodium purification system maintains concentration of impurities below an acceptable level to prevent deterioration of sodium fast reactor (SFR) components. A cold trap is typically monitored with multiple thermal hydraulic sensors. Timely detection of incipient anomalies in cold trap operation is important for efficient SFR operation and maintenance. Previous work developed a deep learning long short-term memory (LSTM) autoencoder for loss-of-coolant type anomaly detection in cold trap of the liquid sodium purification system at the Mechanisms Engineering Test Loop (METL) thermal hydraulic facility at Argonne National Laboratory. We found that relative delays in response time for multi-modal sensor monitoring systems affect anomaly detection time and certainty. We have developed a novel machine learning (ML) method to estimate sensor response delays in detection of signals related to anomaly events, and to use this information to augment the data to improve detection time. The anomaly signal is detected by establishing a threshold using the density distribution of the loss for the training data. Relative sensor delays were determined during testing by finding the times when the loss of each sensor rises above their respective threshold values. The time delays were then used for synchronization of the data. The augmented data was fed back to the LSTM autoencoder to detect the anomaly using sensor-averaged loss. A parametric study was conducted, in which the anomaly was gradually reduced until the signal-to-noise ratios (SNRs) approached unity. Results indicate that synchronization improves anomaly detection, especially for lower SNR anomalies.

The High Temperature Test Facility is an integral test facility located at Oregon State University, modeled after the Modular High Temperature Gas Reactor. It is designed to provide benchmark data for phenomena such as lower plenum mixing, depressurized conduction cooldown, pressurized conduction cooldown, and normal operational conditions. Numerous sensors are installed throughout the facility to measure variables like temperature, pressure, and mass flowrate, with data recorded at 2 Hz frequency. Several methods are under study for field reconstruction in online monitoring applications. One method that is promising with sequential data, but not well-studied in nuclear engineering is Recurrent Neural Networks (RNN). This study focuses on developing data-driven RNN models,specifically gated recurrent units (GRU) and long short-term memory (LSTM), to predict sensor outputs at various locations within HTTF. The models are trained on data from one subset of sensors and applied to predict the outputs of similar sensors - this was done 400 times, with 200 permutations of LSTM and GRU models each. Mean absolute error (MAE) was used as a performance metric to evaluate the predictions. It was found that 71 of the sensors can be used to train LSTM and GRU models, which can then predict the data of the other 71 sensors very well. The MAE of the predictions ranged from 0.28°C to 4.41°C for all models and permutations. Generally, the LSTM models have a higher accuracy relative to the GRU models with overall (average MAE value of 0.721°C for LSTM as opposed to 0.788°C for GRU).

The Shallow Recurrent Decoder networks are a novel paradigm recently introduced for state estimation, combining sparse observations with high-dimensional model data. This architecture features important advantages compared to standard data-driven methods including: the ability to use only three sensors (even randomly selected) for reconstructing the entire dynamics of a physical system; the ability to train on compressed data spanned by a reduced basis; the ability to measure a single field variable (easy to measure) and reconstruct coupled spatio-temporal fields that are not observable and minimal hyper-parameter tuning. This approach has been verified on different test cases within different fields including nuclear reactors, even though an application to a real experimental facility, adopting the employment of in-situ observed quantities, is missing. This work aims to fill this gap by applying the Shallow Recurrent Decoder architecture to the DYNASTY facility, built at Politecnico di Milano, which studies the natural circulation established by internally heated fluids for Generation IV applications, especially in the case of Circulating Fuel reactors. The RELAP5 code is used to generate the high-fidelity data, and temperature measurements extracted by the facility are used as input for the state estimation. The results of this work will provide a validation of the Shallow Recurrent Decoder architecture to engineering systems, showing the capabilities of this approach to provide and accurate state estimation.

Advanced reactor and fuel designs could be crucial in decarbonizing our energy portfolio. However, their development and implementation come with specific challenges, often related to the novelty of such designs, that must be addressed to ensure that such systems operate safely, reliably, and economically viable. Strategies like the preventive maintenance of such systems can support achieving these goals by potentially reducing the maintenance and operation costs while preserving the safety and reliability of such systems.

However, the preventive maintenance of nuclear reactors may rely on real-time monitoring of some physical properties of such systems, which can be challenging. Many probe designs cannot withstand the reactor's extreme conditions (e.g., temperature, radiation). In this context, physics-informed Convolutional Neural Networks (CNNs) offer a promising non-intrusive alternative for reconstructing physical fields, such as temperature and stress distributions, using minimal sensor data. This work presents the integration of machine-learning-aided tools with coupled thermomechanical and thermal-hydraulic simulations to assess the behavior of fuel rods during both steady-state and accident scenarios.

To train our CNN, we leveraged the capabilities of the MOOSE framework to build computational models representing the fuel rod thermomechanical behavior during steady-state operation and its response during a transient situation, such as an accident condition. These models were used to build the necessary datasets to train and test the prediction performed by the CNN architecture. These efforts provide a foundation for real-time monitoring and enhanced safety assessments of advanced reactor designs, addressing challenges in operational efficiency and accident management.

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.