Conference Agenda

Advanced nuclear fuels—such as Accident-Tolerant Fuels (ATF), helical fuel, and Transformational Challenge Reactor (TCR) fuel—have underscored the need for high-fidelity multi-physics fuel performance analysis. To address this demand, the BEEs code (developed by the XJTU-NuTheL research group) offers a multi-physics, multi-dimensional simulation framework tailored for advanced nuclear fuel systems.

By integrating high-fidelity finite element models with advanced coupling strategies, BEEs tackles critical challenges in fuel performance evaluations under diverse operational scenarios. The code features comprehensive models for diverse fuel types, including UO₂-Zircaloy rod fuel, TRISO-coated particle fuel, annular fuel, and plate-type fuel, incorporating thermal-mechanical behavior, irradiation effects (creep, swelling, fission gas release). Validation studies demonstrate good agreement with experimental data and benchmark cases for predictions of temperature, stress, and deformation under normal operation, Loss-of-Coolant Accident (LOCA), and Reactivity-Initiated Accident (RIA) scenarios. Meanwhile, both discontinuous Galerkin and high-order implicit time-stepping methods have been introduced for one-dimensional coolant channel modeling with enhanced accuracy. Application highlights span fuel performance evaluations for ATF, annular fuel, gas-cooled reactor fuels, and plate-type fuel, alongside multi-physics coupled simulations of typical PWR primary circuits and plate-type fuel assemblies.

This work establishes BEEs as a promising code for serving as both a fuel design tool and an independent performance evaluation platform for solid rod fuel, annular fuel, plate fuel, and TRISO.

Flow-induced vibrations (FIV) of key nuclear reactor components, such as fuel rods and steam generator (SG) tubes, may lead to wear and damage. Hence understanding and being able to predict the vibrational behavior of these components is crucial to mitigating any potential risks and preventing undesired outages. Fuel rods primarily vibrate due to the turbulent axial flow, requiring generally scale-resolving models to properly study their vibrations numerically. SG tubes on the other hand are exposed to cross-flow, with vibrations being a result of a combination of turbulence-induced and vortex-induced vibrations, possibly resulting in fluid-elastic instability. This cross-flow nature of the problem may make it possible to study it using computationally less expensive numerical techniques, such as those based on the Unsteady Reynolds-Averaged Navier-Stokes (URANS) approach or hybrid turbulence models. The current paper attempts to give an overview of where we are in terms of our understanding of FIV of a multiple tube configuration in cross-flow and how well these problems can be modelled using medium-resolution numerical approaches. This is done by first considering two well studied problems, being the numerical benchmark of Turek & Hron of a flexible flap attached to a fixed cylinder, and a single cylinder in cross-flow. The former allows one to validate properly the FSI framework used to study cylinders subjected to cross-flow, while the latter serves as a canonical problem fundamental to understanding tube bundles in cross-flow. Following these two cases, two-cylinder systems, with cylinders positioned either inline or side-by-side, and tube bundles are discussed. In general, a lot of data coming from experiments is available for all these cases, allowing one to validate and study them numerically in detail. Also, medium-resolution simulations do provide reasonable predictions for single and two-cylinder configurations, but struggle to recover vibration amplitudes in the lock-in regime. For tube bundles though, the amplitudes are generally overpredicted. This may be caused by a lack of turbulence that is actually resolved, although more detailed benchmark data is needed to further investigate this.

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

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

Improving thermal-hydraulic performance is a major goal for many applications since fluid flow is essential to many natural and artificial systems. This study focuses on assessing thermal and fluid flow performance in a corrugated mini channel, which has a distinct separation and mixing zone arrangement that influences its thermal-hydraulic behavior. To investigate how various geometric parameters affect this channel's hydraulic performance, experiments and CFD simulations were carried out. Using water as the working fluid and volumetric flow rates ranging from 1 to 7 L/min, increasing in increments of 0 to 1 L/min, an experimental investigation was carried out. The Reynolds numbers for these flow rates ranged from 1000 to 4000. The study also explores the effect of the mixing-to-separation-zone length ratio (Lm/Ls) on hydraulic operations. A crucial metric for evaluating hydraulic performance, the friction factor, and Lm/Ls are clearly correlated in the experimental results. This experimental result had a maximum deviation of 5% from the numerical calculation. Consequently, a power law-based novel correlation with a variance of less than 5% is suggested to forecast the friction factor and heat transfer. This emphasizes how the Reynolds number and geometric parameters both affect the friction factor, a crucial hydraulic performance metric.

The accurate prediction of two-phase flow void fraction and wall superheat under pressurized conditions is crucial for understanding reactor safety margins. Existing void fraction and flow boiling heat transfer models used in numerical simulations exhibit significant uncertainty, limiting their accuracy in two-phase CFD simulations. This paper presents high-resolution vertical upward flow boiling experimental data from the PCHT test facility at the University of Michigan, using a gamma densitometer and X-ray imaging system. Experimental results are compared with established one-dimensional models to validate their applicability and identify limitations. The Saha and Zuber correlation is used to predict the thermodynamic equilibrium quality at the net vapor generation point. The slip ratio model proposed by Chisholm, Thom, Zivi, Cai, Lockhart, and Martinelli was used to estimate the void fraction and heat transfer coefficient in sub-cooled flow boiling conditions.

Hence, based on the existing correlations from previous research and experiment data from the PCHT test facility, a new one-dimensional model is proposed to better predict the void fraction and wall heat transfer coefficient under the vertical, upward, sub-cooled flow boiling conditions.

This study presents an innovative use of boiling acoustics techniques to examine the cooling system of the Isotope Production Facility (IPF) at Los Alamos National Laboratory (LANL). In the IPF target station, multiple stacked targets are arranged with a series of water channels interspersed in between, leveraging forced convection for cooling. During operation, the rastered high-energy proton beam can initiate various boiling regions from subcooled boiling to critical heat flux. Identifying these boiling characteristics is challenging because of the extreme radiation environment. For that, a prototypical facility setup with a transparent window is utilized for the visualization of boiling phenomena. In this paper, we employ a hydrophone and high-speed video camera to capture acoustic signals and images indicative of various boiling phenomena. By applying signal processing techniques such as Fast Fourier Transform (FFT) and Short-Time Fourier Transform (STFT), we aim to discern distinct boiling behaviors from the hydrophone data. The insights gained from this analysis will guide the installation of hydrophones within IPF, allowing real-time monitoring to prevent boiling crisis by adjusting operational parameters such as beam intensity. While this methodology is tailored for IPF, its implications extend to other systems where boiling dynamics are critical, particularly in the nuclear industry and research sectors. This research enhances our understanding of thermal hydraulics and heat transfer in isotope production facilities and contributes to the safety and efficiency of nuclear systems.

This study investigates the nucleation site density (NSD) on nuclear fuel cladding materials under flow boiling conditions in an annular gap geometry. Experiments were conducted at the Karlsruhe Institute of Technology (KIT) COSMOS-L facility using three cladding samples: uncoated Zircaloy-4, and two physical vapor deposition (PVD)-coated variants, CrN and Cr. The test section comprised a 9.5 mm diameter cladding heated over a 330 mm length, with data collection focused on a 25 mm segment near the outlet. Measurements were performed at 300 kPa outlet pressure, approx 500 kg/m^2/s mass flux, and an 85°C inlet temperature, with variable heat flux. High-speed videography captured bubble dynamics, and nucleation sites were identified using an in-house KIT code that tracks brightness changes in individual frames to calculate the frequency and spatial distribution of departing bubbles. To distinguish true nucleation sites from passing bubbles, noise, and non-uniform illumination, an adaptive filter based on proper orthogonal decomposition was implemented. The NSD comparison across the three samples revealed observable variations, which require further evaluation under different flow and heat flux conditions to better understand surface modification effects on boiling behavior.

An experimental setup has been designed and manufactured at the Royal Institute of Technology (KTH) to investigate the internal structure of boiling two-phase water flow under prototypical Light Water Reactor core conditions, including those relevant to PWR, BWR and SMR designs. The setup is based on the High-pressure WAter Test (HWAT) loop, designed for 25 MPa pressure, 1 kg/s water mass flow rate and 1 MW thermal power. The facility has been updated with a new test section and advanced instrumentation systems to enable measurements under steady-state and transient operations. This novel experimental setup allows for the first-time measurements of radial distributions of local two-phase flow parameters under high-pressure LWR core conditions. The resulting data is intended to enhance the fundamental understanding of boiling two-phase flow phenomena, contribute to the development of closure laws and support the validation of computational codes. The paper presents the loop design, the updated instrumentation with associated uncertainties, and data post-processing methods (including the derivation of dispersed phase length scales). Results from commissioning tests, such as heat balance tests and single-phase tests, are presented. Examples of high-pressure boiling two-phase flow measurements are presented and discussed. Fundamental behavior and associated key parameters, including radial distributions of void fraction, mixture velocity, interfacial length scales and polydispersed characteristics, are identified and quantified.

Natural circulation flow is widely adopted in Small Modular Reactors (SMR) to simplify system design and achieve passive heat removal. The characteristics of low-velocity natural circulation flow are affected by both forced and natural convection. Since the amount of natural circulation flow is dependent on pressure losses, such as the wall friction, understanding this mixed convection flow is essential for the system design and safety evaluation of SMR. However, most of the previous studies were conducted using air as a working fluid or under low-temperature water conditions. Therefore, this experimental study investigated the wall friction factor in high-temperature water flows. The wall friction factor was measured at a vertical annular channel under natural circulation flow conditions. Experiments were performed under Re of 690-5,020, and Gr of 10⁵-5.5×10⁷ at the heated channel. The gaps of the concentric annular channels were 2.9, 5, and 7 mm, respectively. Evaluation of the existing model showed that the forced convection wall friction model underpredicts the present experimental data under low-velocity and larger gap conditions. Under these conditions, secondary flow within the channel prevented development of flow. Accordingly, this caused continuous changes in the velocity profile, increased viscous dissipation, and greater pressure loss. To predict accurately the increased wall friction factor in mixed convection flow, a new model was developed based on the present experimental data. The model accounts for secondary flow induced by buoyancy and radial temperature gradients within the channel. The developed model demonstrated good prediction under low-velocity mixed convection flow conditions.

In this study, we proposed scaling criteria for wall boundary conditions of thin plates with volumetric heat generation based on analyses of free convection-conduction conjugate heat transfer. Unlike uniform wall temperature(UWT) or uniform heat flux(UHF) boundaries, nuclear fuels typically involve volumetric heat generation. To examine the effect of conjugate heat transfer, the parameter for scaling criteria of wall thermal boundary conditions was analytically derived using the perturbation method. To quantify this parameter, the governing equations were numerically solved using the Runge-Kutta method for free convective flow and the finite volume method for solid conduction. The results showed that when the axial conduction-to-convection ratio—defined as half the plate thickness divided by the product of the modified Biot number and plate length—is greater than 0.5, the solution converges to the UWT solution. Conversely, when this ratio is less than 0.01, the solution aligns more closely with the UHF solution. For plates with the same volumetric heat generation, the peak temperature is highest under UHF condition and lowest under UWT condition. Therefore, the scaling criteria for the wall boundary condition proposed in this study can make a significant contribution to the thermal design of nuclear fuels. Furthermore, the scaling criteria were validated against experimental data for both air-cooled and water-cooled free convective plates with volumetric heat generation. According to the scaling criteria proposed in this study, the air-cooled test data were closer to the UWT condition, whereas the water-cooled test data were closer to the UHF condition.

Natural convection in an inclined channel with downward facing heater were investigate to validate the performace of PECS(passive ex-vessel corium retaining and cooling system), the Korean core catcher developed for the exporting APR1000 reactor. PECS has inclined channels under the structure , and the channels are heated from the top surface when the molten corium drops on the structure by a severe accident. The VPEX(Variable PECS experimental facility)was designed and built to examine the phenomena in the PECS channel experimentally. The VPEX has the heating block made of the carbon-steel with stainless steel coating, which is the same as the PECS structure. And the heat flux distribution over the channel were given by CFD calculation considering the corium behavior on the PECS. The tests were performed for various parameters such as the heat flux, the inlet subcooling, the channel shape, and the pressure. The results shows that the PECS has sufficient cooling capability even with the 175% of the expected heat flux. Also, the behavior of natural convection in the PECS channel were calculated using 1-D code, NCir, and the results were compared with the tests. The calculation results vary by the two-phase friction model and the void fraction model, however, corresponds to the experiments well for certain models.

This study investigates heat transfer enhancement in two-phase natural circulation loops, focusing on the role of superficial velocity in Small Modular Reactors (SMRs). Natural circulation, driven by buoyancy forces, is a key passive cooling mechanism in SMRs, where compact designs necessitate efficient heat removal. Using the MARS-KS code, this research simulates flow behavior, void fraction, mass flow rate, and heat transfer coefficients under two-phase conditions, specifically analyzing the NuScale SMR design, reduced to 1/10th of the original size for experimental feasibility. The study examines the impact of the minimum and maximum superficial velocity ranges, representing variations in the cooling channel gap. The results demonstrate that optimizing superficial velocity enhances thermal efficiency by improving convective boiling and phase change dynamics, while maintaining system stability. Higher velocities lead to better heat transfer performance in the evaporator, riser, and condenser, with the increased flow velocity fostering more efficient heat dissipation. These findings indicate that controlling the cooling channel gap can optimize flow velocity and, consequently, heat transfer. This research provides a foundation for future experimental studies on cooling channel height control, which will further investigate the influence of gap adjustments on heat transfer. The results contribute to the development of more efficient and reliable passive cooling strategies in SMRs, advancing reactor safety, performance, and sustainability. By optimizing natural circulation and refining reactor designs, this study supports the ongoing efforts to enhance the safety, efficiency, and long-term stability of next-generation nuclear energy systems.

External reactor vessel cooling (ERVC) plays a vital role in preventing failure. Nucleate boiling on the curved lower head of the RPV is a key heat removal mechanism. While current models focus on coolability limits, early cooling through nucleate boiling is equally important. Nucleate boiling models rely on accurate heat flux partitioning, with single-phase heat transfer being a key contributor. However, the correlations used in system analysis codes are not suited to the curved geometry of the RPV's lower head. Understanding single-phase heat transfer on downward-facing curved surfaces is essential for developing accurate nucleate boiling models. This study addresses this gap by developing heat transfer correlations for single-phase flow on curved, downward-facing surfaces under constant heat flux. Experimental measurements and CFD simulations were used to develop two correlations: one for forced convection (Ri < 0.1) and another for mixed convection (0.1 < Ri < 10), within the ranges 1,000 < Re < 13,000, 2.56 < Pr < 4.36, and 0.001 < Ri < 10. The study highlights that in forced convection, curvature affects boundary layer development and heat transfer. For mixed convection, buoyancy effects are captured through the Buoyancy number (Bo), with correlations showing how buoyancy transitions from impairment to enhancement of heat transfer. These findings are vital for improving system codes used in nuclear safety analysis, particularly for predicting heat transfer during ERVC and ensuring the integrity of the RPV in severe accident conditions.

In molten-salt-fuelled reactor systems, the fluid may experience substantial volumetric heat generation in addition to heat removal from surrounding structures. To quantify these effects, we investigate developed channel flow with internal heating using a systematic multi-scale approach comprising Direct Numerical Simulation (DNS), Large Eddy Simulation (LES), and a semi-analytical solver (SAS). First, DNS and LES are compared in a turbulent parallel-plate configuration at different Prandtl and Reynolds numbers, demonstrating excellent agreement in flow and thermal fields, with the SAS method showing acceptable accuracy. Building on this benchmarking, the SAS tool is then employed to explore a broad parameter space, offering insights into how internal heat deposition modifies the temperature distribution across Reynolds and Prandtl numbers. Comparisons are also drawn against the canonical wall-heating scenario, revealing that volumetric heating often remains a secondary effect in turbulent regimes but can become more pronounced at lower Reynolds numbers, higher Prandtl numbers, or when nearly all heat is deposited in the fluid. These findings establish guidelines for reduced-order modeling in liquid-fuel reactor analyses and highlight conditions under which internal heating warrants particular attention. The paper concludes by outlining ongoing and future research directions, including refinements for variable fluid properties and complex geometry extensions.

Moltex Energy is developing the technology and design for a stable salt reactor-waste burner (SSR-W). It is a static fueled chloride molten salt reactor (MSR) with a fast neutron spectrum and is designed to be fueled with transuranic elements recovered from spent fuel of CANDU and light-water reactors. Argonne, University of Michigan, and Moltex are developing three multi-physics plant digital twins (DT) which leverage advanced computational methods to achieve reactor performance optimization and enable predictive maintenance to reduce the plant operation and maintenance (O&M) costs. DT-1 provides methods for long-term fuel cycle modeling and optimization and provides the operator with an indication of refueling time and position. DT-2 comprises of an integrated system plant model that can be utilized in simulating normal operation as well as in assessing the safety performance of SSR-W during postulated and bounding accident conditions. DT-3 develops a conceptual structural health monitoring strategy for an innovative matrix heat exchanger, which involves machine learning-based classification of distributed temperature sensing for detection and localization of faults. Additionally, a separate effects loop (SEL) is being constructed at Argonne to support the SSR-W in three technical areas: (1) thermal-hydraulic heat transfer in MSR, (2) natural circulation phenomenon pertinent to passive decay heat removal, and (3) redox control in salt to minimize corrosion of structural materials. This paper describes the methodologies implemented in the three DTs and shows some illustrative results and outputs. The paper also describes the SEL, the test campaign and presents some preliminary thermal hydraulic and heat transfer data.

Aqueous homogeneous reactor is a new type of reactor that dissolves nuclear fuel in a liquid. The aqueous homogeneous reactor have significant advantages in extracting medical isotopes such as I-125 and Sr-89 due to their liquid fuel characteristics. However, during the operation of the aqueous homogeneous reactor, water molecules in the fuel aqueous homogeneous will collide with fission fragments, decompose to produce hydrogen and oxygen. Under accident conditions, the rapidly increasing nuclear power of the reactor will exacerbate this phenomenon. The generated hydrogen will accumulate in the gas space of reactor, posing a threat to the structural integrity of the reactor. Therefore, the system design of an aqueous homogenous reactor needs to take the hydrogen elimination into consideration.This article considers a gas recombination system(GRS) , a nitrogen blow system(NBS) and constructs a 200kW aqueous homogeneous reactor model in RELAP5.The impact of the gas Recombination system on the volume fraction of hydrogen produced by the aqueous homogeneous reactor in a typical reactivity introduction accident was analyzed under different working conditions. The results indicate that the presence of a gas recombination system can significantly reduce the volume fraction of hydrogen during a reactive accident in a aqueous homogeneous reactor. The hydrogen volume fraction can be guaranteed to be less than 4% during the accident , which make sure meets the requirement of hydrogen risk guideline. This study contributes to the design and construction of medical aqueous homogeneous reactor.

Due to the high-temperature operation expected from Molten Salt Reactors (MSRs), thermal radiation can significantly influence the thermal-hydraulic evaluation of these systems, and the intricate coupling of multiple physical phenomena in this kind of reactor necessitates the development of high-fidelity simulation codes. This work presents a new library, developed in OpenFOAM using C++ object-oriented programming, to model thermal radiation coupled with fluid mechanics and other physical phenomena. This library employs the SP₃ method to solve the Radiative Transfer Equation (RTE) and is compatible with all OpenFOAM sub-models for absorption, emission, and scattering, thanks to its inheritance from the default radiationModel library. Additionally, an innovative boundary condition is implemented to model metals such as Stainless Steel and Hastelloy, which are commonly used in the reflector of MSRs. This code can accurately calculate the power density distribution within the system through the strong coupling between neutronics and thermal-hydraulics calculations. This work considers two case studies: the 2D axisymmetric EVOL geometry and the 3D one-sixteenth MSFR geometry. In both cases, the effects of the thermal radiation on the temperature fields are evaluated. Additionally, the effects of temperature, fluid flow type (laminar or transient), and radiation properties (emission and absorption coefficients) on incident radiation are analyzed.

Kairos power is developing a pebble fuel based molten salt reactor. As a part of development, pressure drop through the pebble bed is tested, and a converging and diverging conical part of the core is experimentally studied in current scope. Test facility is scaled down with pebble’s Reynolds number. Test results are compared to KTA correlation, which is widely used correlation in cylindrical pebble bed pressure drop. Since KTA correlation is based on cylindrical geometry, parameters are taken at inlet or outlet of each truncated conical interval. For converging cone, there are four measurement intervals along flow direction. Most of the test results show good accordance with KTA correlation based on parameters from outlet of the truncated cone. However, the last interval, which has the smallest cross section, KTA underestimates the pressure drop. The difference between KTA and test at the last interval increases with Rep. For diverging cone, there are three measurement intervals along the cone. For the middle interval, KTA and test result match well. However, for the other two intervals, KTA shows significant overestimation. Considering manifold like geometry of the two cones and interval-sensitive predictability of KTA, a new correlation is needed.

Simulating nuclear transients with Computational Fluid Dynamics (CFD) poses significant computational challenges due to complex physics and disparate temporal scales. These lead to high computational costs, necessitating advanced techniques for efficiency. This work explores two strategies: multi-rate timestepping with overset grids and reduced-order modeling (ROM).

First, we present an overlapping domain capability within NekRS, a GPU-accelerated spectral element CFD solver. This feature enables independent solution of spatial regions, improving efficiency for large-scale transients in complex geometries. We demonstrate its scalability through the TALL-3D experiment, a benchmark for thermal-hydraulic behavior in liquid metal reactors. Multi-rate timestepping significantly accelerates simulations, addressing a key CFD bottleneck.

Second, we develop ROMs as a computationally efficient alternative for high-fidelity transient simulations. ROMs support digital twin technologies, enabling rapid simulations with high accuracy. Using the NekROM framework, we implement Proper Orthogonal Decomposition (POD)-based ROMs to tackle challenges in nuclear modeling, including (1) thermal striping, which induces fluctuating thermal stresses affecting structural integrity, and (2) molten salt reactor (MSR) modeling, an emerging reactor technology. Both cases involve long time scales, benefiting from ROM acceleration. Our results show POD-based ROMs achieve significant computational speedup while maintaining essential accuracy.

These advancements in multi-rate time-stepping and ROMs represent a major step forward in transient simulation. By improving computational feasibility, they enable more efficient and accurate simulations of complex nuclear systems, enhancing reactor design, safety analysis, and operational decision-making.

In the realm of computational fluid dynamics (CFD), the high computational cost has always been a significant challenge, particularly in fields like nuclear safety analysis where complex flow problems are common. However, scientific machine learning (SciML) has demonstrated its effectiveness in solving real-world problems, including those in CFD. In recent studies, the issue of residual divergence in long-term simulations when using a single training approach has been identified. Our goal is to develop a flexible framework to optimize the state-of-the-art hybrid ML algorithm; residual based physics informed transfer learning (RePIT) developed by J. Jeon et al. This is a promising technique which has accelerated the simulation and ensures long-term stability using neural networks. However, its performance was only demonstrated on one network architecture known as finite volume method network (FVMN), also manual intervention was involved while switching between ML and CFD computation. Our projection is that we can further reduce the computational time by automating the switching process and also adding several network benchmarks where we could be able to implement state-of-the-art neural network architectures and choose the best performing one. In this work, we verified (1) the integrity of the framework by using the FVMN and compared the results with the original case study and (2) that various ML models could be loaded in the RePIT framework. In particular, DeepONet-RePIT shows the best acceleration performance. We believe that this hybrid approach is the most practical SciML utilization for robust CFD acceleration.

Typical prismatic core High Temperature Gas Reactor (HTGR) designs feature many individual coolant channels that are long in the axial direction and circular in cross section. When predicting fuel temperatures during initial concept design iterations, it is appropriate to model the steady, single-phase flow and heat transfer within each of these coolant channels using simple correlations. The challenge is then providing appropriate thermal boundary conditions along the walls of every individual channel in the core (e.g. accounting for power shapes). For complex core layouts, conjugate 3D CFD models of the entire assembly could be used to address this at the expense of significant computation time.

In this work we demonstrate a hybrid approach that uses a fluid-as-a-solid approximation to enable 3D simulations of entire core assemblies at reduced computational cost. This eliminates the need to solve the Navier-Stokes equations in the fluid, while leveraging existing functionality available within the CFD code STAR‑CCM+ to minimise implementation time and enable rapid design studies.

The approximation involves the fluid in each channel being modelled as having a suitable anisotropic effective thermal conductivity. This study considers gas flows within geometries relevant to prismatic HTGR cores. Predicted temperatures are compared between baseline CFD simulations and simulations employing the fluid-as-a-solid approximation. The fluid-as-a-solid approximation predicts similar temperature profiles to the baseline simulations. A reduction in computation time of around two orders of magnitude was achieved. This approach is expected to be valuable to preliminary concept designers, wherein rapid turnarounds are desirable across a range of designs.

In advanced reactor designs, dual-cooled annular fuel assembly has attracted significant attention due to its unique thermohydraulic characteristics. However, its complex structure poses challenges for the traditional analysis methods. In this paper, a Physics-Informed Coarse-Mesh method is proposed and applied to the analysis of annular fuel assembly. Given that annular fuel consists of internal and external channels, coarse meshes are employed to capture the primary geometric features, thereby limiting computational costs. Widely validated empirical correlations are used to correct wall friction and heat transfer, ensuring simulation accuracy. By developing a conjugate heat transfer method and a one-platform multi-scale coupling strategy with the fine-mesh CFD method, the issues of flow and heat distribution within annular fuel assembly are resolved. Based on the design parameters of the OPR-1000 reactor, a comparison of the thermohydraulic characteristics between cylindrical and annular fuel assemblies is conducted. The results show that annular fuel assembly exhibits lower central fuel temperature and higher pressure drop. The flow rate between different internal channels remains consistent. Furthermore, comparisons with existing programs demonstrate that this method can accurately simulate the flow and heat transfer characteristics of annular fuel assemblies, providing robust support for the design and optimization of advanced nuclear reactors.

The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) is a research reactor capable of delivering large-volume, high-flux thermal neutron irradiation in a realistic environment. Its design enables comprehensive studies on the effects of intense radiation on reactor materials and fuels. To ensure these experiments are conducted under specific conditions while maintaining safety standards, rigorous programmatic and safety analyses are required. These analyses typically consider coupled physics such as thermal-hydraulics, neutronics, structural analysis, and fuel performance. In this work, the Multiphysics Object Oriented Simulation Environment (MOOSE) framework and MOOSE-based applications are used for developing coupled multiphysics simulations for the ATR core. Regarding thermal-hydraulics analyses, traditional high-fidelity computational fluid dynamics (CFD) are often computationally expensive and the available codes do not have the physics models required for the simulation of the other aspects necessary for reactor analysis. This study leverages the coarse-mesh CFD capabilities in Pronghorn for conducting coupled thermal and neutronics analyses of the ATR core using a simplified porous-medium approach. This method homogenizes solid and fluid regions, enabling streamlined geometry and accelerated simulation times. This paper aims to: i) develop a 3D model of the ATR using publicly available data; ii) create a corresponding coarse-mesh CFD model; iii) verify simulation results against benchmark calculations; and iv) evaluate the use of the porous-media methodology for simulating the ATR. The results indicate that the coarse-mesh CFD capabilities provide accurate predictions for the temperature difference and pressure drop at the core, and fuel temperature distribution of the ATR, with improved run-time.

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

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

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

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

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

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

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

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

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

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

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

Direct Air Capture (DAC) is a technology that separates and captures CO₂ contained in trace amounts in the atmosphere. It is the only existing CO₂ capture process with a negative net emission and is receiving attention as an active CO₂ removal technology (Carbon Dioxide Removal (CDR)) to achieve net zero. Currently, DAC has low technological maturity overall, and the large amount of air intake requirement and the large amount of heat energy consumption for regeneration are pointed out as major bottlenecks in technology commercialization. Linking with nuclear power, a carbon-free power/heat energy source, is one of the effective strategies to resolve the technological bottleneck of DAC, but the research and development for this is still in the basic development stage. In this paper, at first, it is explored the concept development research of the nuclear power-DAC combined system currently in progress and evaluate the development level. In addition, it will review the technical feasibility of the DAC system, development of the system, the current level of technological development, and propose technology candidates for linking with nuclear power.

In the Republic of Korea, the development of the innovative Small Modular Reactor (i-SMR) is ongoing. The i-SMR will be equipped with the following three passive safety systems to replace the active safety systems of existing commercial nuclear power plants: Passive Emergency Core Cooling System (PECCS), Passive Auxiliary Feedwater System (PAFS), and Passive Containment Cooling System (PCCS). The PECCS performs the core makeup/cooling function, the PAFS removes residual heat with theby steam generator (SG) cooling, and the PCCS conducts the heat removal from the containment vessel (CV) atmosphere. Since the passive safety system can carry out safety functions by natural forces without continuous power supply or any operator action, it is expected to dramatically improve the safety of nuclear power plants (NPP). In this paper, we present the design status of the passive safety systems in the i-SMR and the performance analysis results using MARS-KS.

This paper extends the evaluation of an adaptive algorithm for estimating decay heat in transient scenarios with steadily decreasing reactor power. The previously introduced approach offers a low-cost alternative with simpler implementation to traditional methods that typically rely on extensive nuclide tracking or standardized procedures. The adaptive algorithm utilizes precomputed decay heat curves from SCRAM scenarios, and enables real-time decay heat estimation during simulation while dynamically adjusting to varying power levels without requiring detailed nuclide tracking.
This study applies the adaptive algorithm to additional non-SCRAM shutdown scenarios in a sodium-cooled fast reactor, demonstrating its flexibility in handling diverse transient conditions. Verification is performed by comparing results with Monte Carlo depletion calculations using Serpent, confirming high accuracy. Furthermore, the model is successfully applied to FFTF LOFWOS Test #13, showing good agreement with experimental decay heat data. These results demonstrate the potential of the adaptive method as an accurate decay heat estimation approach for real-time transient simulations of advanced reactors.

The water-based Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory is a large-scale test facility built to study passive decay heat removal performance of one Reactor Cavity Cooling System (RCCS) concept for advanced nuclear reactors. The inventory fill within the primary water tank is known to have an impact on the natural circulation thermal hydraulics and two-phase phenomena development, including instabilities such as startup oscillations. Six inventory levels, from 20% to 80% initial fill, were examined. The facility was heated from room temperature to an input power corresponding to 2.1 MW_t full-scale (51.6 kW_t NSTF-scale) and operated at saturation conditions for at least four hours, uninterrupted. While minor changes to integral thermal hydraulic characteristics and instabilities were observed, ultimate heat removal performance was not impacted by decreasing inventory. Following the last 20% fill test, a depletion scenario was performed where the facility continued operating at saturation conditions with boil-off until reaching critical levels where natural circulation flow stagnated. Comparisons were made to similar, previous inventory parametric series and depletion tests. The height of the tank inlet has an impact on the development and suppression of the natural circulation instabilities as well as natural circulation stagnation conditions. However, the lower tank inlet resulted in significant increase in available inventory prior to stagnation in a boiloff scenario. Additionally, no short-circuiting effects between the hot and cold legs were observed as a result of the lower tank inlet.

The objective of the present study is to investigate the effects of varying decay heat loads and tank inlet elevation on the performance of a large-scale water-based Reactor Cavity Cooling System (RCCS). The water-based Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne represents a ½ axial scale and 12.5° sector slice of the full-scale Framatome 625 MWt SC-HTGR RCCS concept. A power parametric series with prototypic decay heat loads of 1.4, 1.75, 2.1, and 2.4 MWt (34.4, 43.0, 51.6, and 58.5 kWt scaled) with the lower tank inlet configuration was first discussed. System performance metrics at the two-phase quasi-steady state were compared, mainly the system flow, steam generation rate, system pressure, and liquid and structure temperatures. The effect of the decay heat power level on the boiling front progression was also investigated. It was found that lower decay heat loads would cause delayed boiling front progression. Results were then compared to previous data sets collected at identical testing condition but with chimney pipnig configured at the mid tank elevation configuration. Overall, the tank inlet elevation did not significantly influence the system two-phase quasi-steady-state performance at the studied decay heat levels. However, the tank inlet elevation was found to have significant impacts on the two-phase flow instability characteristics and the boiling front propagation. At the lowest power of 1.4 MWt with the lower tank inlet configuration, boiling never propagated from the tank to the upper chimney, while boiling propagation to the upper chimney was observed in the mid tank inlet configuration.

Digital twins are becoming essential in advanced nuclear reactor technology for real-time monitoring, predictive maintenance, and optimization. They help maximize uptime, predict failures, and enable cost-effective testing. However, most digital twins rely on deterministic models that fail to capture uncertainties in sensor data and modeling. While Bayesian models can quantify uncertainty, they require significant computational resources and do not scale well, making them impractical for real-time applications. This work presents the application of variational Bayesian last layer (VBLL) augmented recurrent neural networks (RNNs) to produce uncertainty-aware models while mitigating the issue of computational cost through variational inference. Variational inference allows the neural network to assign parameters to probability distributions rather than point values, and thus permits sampling of predictions to measure the uncertainty of the model. We apply both deterministic Long Short-Term Memory (LSTM) and VBLL LSTM to time-series sensor data collected from the High Temperature Test Facility (HTTF) at Oregon State University and train the models using data from a depressurized conduction cooldown (DCC) experiment. The deterministic LSTM and VBLL LSTM both achieve impressive predictive capabilities with ( R^2 >) 0.99 when forecasting solid and fluid temperature sensor profiles. Despite a small increase in computational cost, the VBLL LSTM is a promising direction for incorporating model uncertainty for real-time applications.

In recent years, image processing methods for gas-liquid two-phase flow, including computer vision techniques, bubble segmentation, and tracking algorithms, have seen significant development due to high efficiency and accuracy. Nevertheless, obtaining extensive, high-quality two-phase flow images continues to be a time-intensive and costly process. To address this issue, a generative AI architecture called bubbly flow generative adversarial networks (BF-GAN) is developed, designed to generate realistic and high-quality bubbly flow images through physically conditioned inputs, namely superficial gas and liquid velocities.

Initially, 105 sets of two-phase flow experiments under varying conditions are conducted to collect 278,000 bubbly flow images with physical labels of and as training data. A multi-scale loss function of GAN is then developed, incorporating mismatch loss and feature loss to further enhance the generative performance of BF-GAN. The BF-GAN’s results indicate that it has surpassed conventional GAN in generative AI indicators, establishing for the first time a quantitative benchmark in the bubbly flow domain. In terms of image correspondence, BF-GAN and the experimental images exhibit good agreement. Key physical parameters of bubbly flow images generated by the BF-GAN, including void fraction, aspect ratio, Sauter mean diameter, and interfacial area concentration, are extracted and compared with those from experimental images. This comparison validates the accuracy of BF-GAN's two-phase flow parameters with errors ranging between 2.3% and 16.6%. The comparative analysis demonstrates that the BF-GAN is capable of generating realistic and high-quality bubbly flow images for any given and within the research scope, and these images align with physical properties.

Boiling is a process in which heat from a submerged surface is removed through bubble vaporization. Consequently, the primary mode of thermal transport is governed by bubble dynamics. Therefore, one way to extract bubble data for estimating cooling performance is through the use of side-view shadowgraphs. However, linking bubble dynamics to other parameters remains challenging due to irregular bubble shapes and varied visualization setups. Even with neural networks, manual data annotation for initial model training demands significant time and effort, further complicated by instrumental differences such as varying lighting conditions. To overcome such facility limitations, this study utilizes an augmentation method to generate a synthetic bubble swarm dataset by assembling individual synthetic bubbles tailored to the experimental setup. This dataset is used to further train a Mask R-CNN for segmentation tasks to automate accurate bubble region predictions. After training, the model successfully identified bubble boundaries in high-speed camera images, extracted size distributions, and derived bubble trajectories under different heat fluxes. This demonstrates the feasibility of training models on augmented images to accurately segment bubble regions from experiments and to provide reasonable estimations of the sizes and trajectories of bubbles. This method serves as an alternative to traditional boiling heat transfer measurements, especially in setups where direct measurement of bubble statistics is limited or under conditions where conventional methods face material and instrumental constraints.

Heat transfer on boiling surfaces is a transient multi-dimensional process. Gaining an understanding of the dynamics of bubbles on such surfaces is a fundamental step toward an accurate heat partitioning models, which can be employed as best estimate models for LWR analysis. A number of traditional techniques have been developed for measuring and interpreting bubble dynamics, but they are time and computation demanding. Machine learning and computer vision have shown great potential to reduce this data interpretation burden in local feature detection such as bubble dry spot. However, a holistic segmentation of boiling features on heat transfer surfaces has not yet been demonstrated. In this study, we employed a U-NET based convolution neural network to conduct a global segmentation of a boiling surface, which allows for the classification of the dry areas, micro-sublayers, and single-phase areas from high-speed infrared images. For training the model, we obtained the ground truth partitioning from HSV phase detection images. After training, the U-NET model can predict the dry areas, micro-sublayers, and single-phase areas directly from the IR image. This successful demonstration paves the way for further research on predicting heat transfer performance based on IR images.

Thermal-hydraulic simulations of Once-Through Steam Generators (OTSGs) are crucial for operational optimization, real-time monitoring, and the development of digital twins. However, repeated and extensive simulations are computationally expensive. Model Order Reduction (MOR) techniques, such as Proper Orthogonal Decomposition (POD), offer an alternative approach to enhance computational efficiency while maintaining acceptable accuracy. Although POD is widely used for capturing dominant patterns in high-dimensional systems, its robustness within the parameter domain is limited due to its reliance on snapshots and potential inadequacy in representing highly nonlinear or complex systems. In this paper, we propose a novel framework for Parametric POD that integrates Long Short-Term Memory (LSTM) networks to enhance the accuracy and robustness of reduced-order models, and applied it to construct a Reduced-Order Model (ROM) of OTSGs. The framework leverages high-dimensional snapshots generated under varying reactor power levels, reducing them to a few dominant POD modes. Given the time-dependent nature of both the parameter (reactor power) and POD mode coefficients, LSTM is employed to approximate the mapping function between them. The resulting parametric ROM is verified for rapid estimation of primary and secondary side fluid temperatures in OTSGs using RELAP5 simulation results. The ROM achieves a maximum instantaneous fluid temperature deviation of less than 2.5 K (0.5% relative error) and reduces computation time to 1% of that required by RELAP5. This novel approach demonstrates significant potential to address computational challenges posed by numerous simulation inquiries, thereby enhancing the efficiency and applicability of OTSG modeling.

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.

Small Modular Reactors (SMRs), particularly Light Water-SMRs, represent a viable option for near-term nuclear deployment in Europe, building upon established Light Water Reactor technology while incorporating evolutionary design modifications like passive safety systems. While these systems offer advantages such as independence from external power and component minimization, they face potential functional failures due to Natural Circulation issues. Therefore, passive system failures must be addressed in SMR design and safety reviews. Current guidance on passive safety system requirements and failure mode modeling methodologies needs consolidation. Short-term research priorities include reliability analysis of ThermalHydraulic phenomena driving system operation and related Uncertainty Analysis. Building on ENEA's MELCOR input-deck developed within the Horizon Euratom Safety Analysis of SMR with Passive Mitigation Strategies-Severe Accident (SASPAM-SA) project, this work applies the REPAS (Reliability Evaluation of Passive Safety Systems) methodology to the Emergency Heat Removal System (EHRS), a key passive safety feature for decay heat removal in advanced designs. REPAS will help analyze various system states, including low-probability scenarios, to understand EHRS behavior and its plant impact.

The flow characteristics in the upper plenum of molten salt (MSR) and high-temperature gas-cooled reactors (HTGR), during the pressurized conduction cooldown accident (PCC) scenario, are dominated by natural convection jets emitted from the top of the reactor core. Benchmark experiments are necessary to validate the Computational Fluid Dynamics (CFD) codes currently being used to further characterize this phenomenon. This study focuses on providing benchmark data for the upper plenum PCC scenario. The experimental facility is a scaled-down generic model of the upper plenum of MSRs and HTGRs. This study produces velocity field measurement data, via Particle Image Velocimetry (PIV). Data from optical fiber-distributed temperature sensors, thermocouples, and volumetric flow data are additionally used to calculate the conjugate heat transfer characteristics of the plenum and provide boundary conditions for the CFD models. The test matrix consists of isothermal and non-isothermal cases. In the non-isothermal case, fluid flow is driven by natural convection and buoyancy forces while the isothermal case is driven by pump-induced pressure gradients. This paper presents a detailed description of the experimental methods and analysis techniques utilized in this study and the results of multiple isothermal and non-isothermal test cases.

Deep pool reactors are reactors that operate at low pressures and are usually built near residential areas for heating and cooling, so the design of the reactor system requires high intrinsic safety characteristics. The design of non-energetic safety systems for deep pool reactors is characterized by multiple novel safety devices, and the transient operating characteristics of the devices are critical to the safety and stability of the reactors. In this paper, a set of experimental system that can reproduce the non-energetic waste heat export function of the pool reactor is set up, and the transient characteristics of the equipment under the loss-of-flow accident are experimentally investigated. The experimental results verify the on-time response of the relevant non-energetic equipment after the design accident. Extended working condition numerical studies were also carried out using a system analysis program.The experimental results demonstrate that the deep pool reactor can export the waste heat of the reactor through its own non-energetic safety system under a loss-of-flow accident.

Passive safety systems are an economically interesting alternative to conventional active systems, which are also more robust against many external influences, as they do not rely on an external, active drive. Accordingly, they continue to function even if large parts of the plant infrastructure are damaged or unavailable, as was the case in Fukushima Daiichi accident, for example. However, passive heat removal systems in particular pose major challenges for designers and the thermal-hydraulic calculation tools they use. One reason for this is the coupling or feedback between the heat flow that is introduced into the coolant and the mass flow of the coolant through the system, which results from the heat input. In addition, state of the art heat transfer models obviously cannot capture the heat transfer for passive systems precisely enough. As a result, attempts to recalculate experimentally determined heat transfer rates of passive heat removal systems with fluid dynamic codes have sometimes resulted in considerable deviations between experimental and numerical data. This paper presents two new heat transfer models developed specifically for passive systems. It is described how they can help to better calculate and predict the performance of related systems. The models were developed primarily on the basis of experimental data recorded at the COSMEA and NOKO test stands and published by the Helmholz Center Dresden Rossendorf. In addition, the model development was supported by CFD calculations executed to better understand the underlying mechanisms.

In the Small Innovative Helium-Xenon Cooled MObile Nuclear power System (SIMONS), the reactor core is horizontally oriented. To investigate the thermal-hydraulic effects of this configuration, a computational fluid dynamics (CFD) methodology is utilized. The flow and heat transfer dynamics of the core assembly under natural circulation conditions are examined, utilizing the passive gas circulation test loop at Shanghai Jiao Tong University as a reference model. The influence of heating power on mass flow rate, Reynolds number, pressure drop, and Nusselt number is examined, delineating the power range that enables flow self-compensation within the natural circulation system. Moreover, this investigation delved into the distribution patterns of mass flow rate, temperature and heat transfer across various channels under conditions of reduced natural circulation. The findings revealed that with a decrease in mass flow rate, there is a progressive increase in the proportion of flow rate and heat transfer within the lower-positioned channels, And the matching degree between the flow rate and heat exchange of each channel decreases. Furthermore, the study observed a pronounced thermal stratification in the upstream chambers at reduced flow rates, which can be attributed to the more pronounced heating exerted by the bottom of the assembly, coupled with the obstruction of flow in the upper chamber by the buoyant ascent of low-density Helium-Xenon mixture.

The heat transfer mechanisms of flow boiling are still unclear. Recent research shows that evaporation of microlayer contributes to bubble growth in pool boiling. In order to investigate the microlayer heat transfer and dynamics, the 650nm-thick indium-tin-oxide (ITO) film is deposited on a 1mm-thick sapphire substrate to heat the fluid. The thickness of the microlayer underneath the bubble was measured using high-speed laser interferometry (LIF) and the transient temperature distribution on the wall was measured by an infrared (IR) camera. Another camera was used to capture the side bubble image. The three devices worked synchronously. The experiment was conducted at 0.11MPa with deionized water as the fluid, covering heat fluxes of 110-174.4 kW/m², subcooling degrees of 0-11.2 °C, and liquid flow velocity of 0.12-0.27 m/s. The typical bubble behavior was analyzed, including bubble growth, sliding, departure and wall temperature distribution. The inverse heat conduction problem (IHCP) of the boiling surface was solved based on conjugate gradient method (CGM) for wall heat flux partitioning. The wall heat flux partitioning outcome has revealed that the formation and evaporation of microlayer have an important effect on the growth of flow boiling bubbles.

Annular two-phase flow occurs in multiple types of nuclear reactors under normal operating conditions and accident scenarios. Annular flow typically features high flow quality and relatively thin liquid film around the fuel elements in nuclear reactors. This flow structure has significant safety implications since the liquid film could break into rivulets due to vaporization and entrainment. Important parameters to characterize annular flow consist of liquid film thickness, wave velocity on the surface of the liquid film, the size and velocities of entrained liquid droplets, and entrainment fraction and rate. This study contributes to existing database of annular flow with both liquid film and droplet measurements conducted with an air-water test facility. The experimental test section consists of a vertical pipe with an inner diameter of 9.525 mm and a length of 2.9 meters, as well as two measurement ports designed for the measurement of liquid film thickness and surface wave velocity by two sets of parallel-wire conductance probes placed at each port. To capture liquid droplet size and velocity, two high-speed cameras are used to capture the droplet field as they exit the test section outlet after extraction of liquid film. The annular flow testing matrix consists of an array of inlet conditions with superficial gas velocity ranging from 7.80 m/s to 34.91 m/s and superficial liquid velocity ranging from 0.09 m/s to 0.44 m/s, spanning the entrainment and non-entrainment annular flow regimes. Acquired data has been used to validate and improve existing correlations or closure models for annular flow.

The experimental characterization of thin liquid films in annular flow regime is central in many engineering applications, ranging from chemical industry to refrigeration systems, and in particular to cooling of light water nuclear reactors, where it is crucial for safety thermal analysis of boiling water reactors and for validation of system codes as well as CFD codes. However, characterization of thin films in annular flow is particularly challenging given the flow turbulence and the high non-linearity of the free-surface behaviour. Particularly, film thickness and wave characteristics are very challenging to be measured, with many correlations from the literature showing substantial offsets in predicting the same quantities under seemingly close boundary conditions. In this paper, results of simultaneous measurements of vertical upward annular flow film in an adiabatic test section is presented using three different techniques. These consist in: a) total internal reflection method, providing highly resolved local thickness measurements; b) X-ray attenuation method, providing interfacial topology and void fraction that can be converted into thickness information; and c) a conductivity film sensor providing high speed thickness and wave information with a spatial resolution of 2 mm. By performing independent calibrations, the three techniques are cross-validated within the corresponding uncertainties. To the authors’ knowledge, this is the first time that the three measurement techniques for film thickness are combined, thus constituting a unique benchmark. The three techniques complement each other and provide highly reliable measurements of annular flows, which are also compared to existing correlations available in the open literature.

Safety evaluation of offshore floating nuclear plants requires boiling two-phase flow data under heaving conditions with long wave period (10–20s). Void feedback effect necessitates investigating the mutual influence of oscillating coolant flow and thermal power.

This study conducted forced convection boiling experiments with sinusoidally oscillating coolant flow velocity and thermal power in an annular double wall channel with an electrically heated rod at atmospheric pressure. The inner diameter of flow channel was 20 mm, and outer diameter of the heater rod was 10 mm, with a heated length of 2000 mm. The axial power profile of heater rod was uniform. The time-averaged inlet flow velocity and linear power density were maintained 2 m/s and 15.7 kW/m respectively. Experiments considered no oscillation and 0.05 and 0.1 Hz sinusoidally oscillating inlet flow velocities and thermal powers with ±15% zero-peak amplitude. The distributions of two-phase flow parameters and flow regimes were identified using an optical void probe with radial traverse at two heights and stereo high-speed camera. An optical fiber sensor with a sheath tube was mounted on the heater rod surface to obtain the axial temperature distribution and capture heat transfer characteristics.

The void fraction oscillated corresponding to the inlet oscillations, significantly at the top of the heated area and slightly in the middle, and its amplitude was affected by the oscillation frequency. This behavior may be attributed to the oscillatory acceleration of the fluid flow. The radial distribution showed less bubble formation in the middle rather than at the top.

The experimental characterization of thin liquid films in multiphase annular flows is particularly relevant to the safety analysis of the cooling channels of light water reactors and for validation of best-estimate system codes as well as CFD codes. The total internal reflection method (TIRM) is an optical method known for decades for being able to non-intrusively measure film thickness of a wide range of fluids flowing over a transparent wall. This measurement is performed by recording with a camera the reflected circular pattern of a laser beam pointed to the flow. The transparent wall is often curved (such as in a pipe), which leads to a potential loss of information, since part of the reflected pattern has to be discarded from each frame because of optical distortions from the curved wall surface. This is also the case for the TIRM experiments performed in our laboratory on adiabatic vertical upward annular flows in a circular section pipe. However, in this work, an innovative approach is developed to use the information on the shape of the distortion rather than discarding it, thus maximizing the value of the measurement and improving accuracy. This is achieved thanks to a thorough data analysis backed up by a previously validated optical ray tracing simulation that replicates our TIRM experiments including the distorted patterns. From the combination of simulation and experiments, new insights are gained into the potential and the limits of standard TIRM film thickness measurements applied to curved pipes.

The study of boiling flow is crucial for designing safety features of chemical and nuclear plants. The boiling behavior and bubble dynamics depend on various system parameters such as pressure, subcooling, mass flux, heat flux, and flow geometry. While a previous flow boiling experiment by the authors’ laboratory has revealed valuable parametric effects in a decent operational range, its setup alone is insufficient to reveal the influence from the channel width. A new dataset of upward flow boiling is therefore collected in a narrower annular test section. Compared to the previous configuration, this new channel has a larger inner diameter of 25.4 mm and the same outer diameter of 38.10 mm, with a heated inner rod of 3-m long. Multi-sensor conductivity probes are adopted measuring void fraction, gas velocity, and interfacial area concentration following a two-group description. Traversing mechanisms are employed allowing the probes to scan across the flow area, and a dedicated sensor pattern is designed and validated to minimize near-wall blind zones for the narrow channel. In addition, high-speed visualization is conducted recording axial flow evolution, and the Onset of Nucleate Boiling and Point of Net Vapor Generation are identified and recorded. Parametric studies are also presented investigating the flow field dependence on systematic boundary conditions. This work presents valuable new experimental data on narrow-channel flow boiling.

Due to the fuel self-shielding effect, reactor irradiation, and component arrangement, plate-type fuel elements exhibit significant transverse non-uniformity in heat generation. Consequently, the internal thermal-hydraulic characteristics of coolant channels may differ from those of conventional channels. To address this, experimental studies on single-phase flow and heat transfer in narrow rectangular channels with transverse non-uniform heating were performed using PIV visualization technology. Analysis of the flow and heat transfer characteristics in narrow rectangular channels revealed the following: Time-averaged velocity field analysis demonstrated that low-velocity regions in laminar flow regimes are more prominent compared to turbulent flow regimes, while both laminar and turbulent regimes exhibit distinct stratification phenomena in transitional flow regions. In laminar regimes, non-uniform heating reduces the nominal boundary layer thickness , with no observed correlation between velocity gradient and heating power magnitude. In turbulent regimes, non-uniform heating increases the nominal boundary layer thickness, and a decreasing trend in velocity gradient is observed with increasing non-uniform heating power.

Under the conditions of a reactor accident, heat transfer at the wall can be hindered, leading to the risk of boiling crisis. The calculation of the near-wall void fraction is crucial for predicting the boiling crisis. In narrow rectangular channels, large-scale interfaces exist near the wall due to geometric constraints. This paper is based on an improved two-phase multi-scale interface model that considers the interfacial transfer of concentration, momentum, heat, and mass for bubbles of different scales. The model is embedded within the Eulerian two-fluid model in Fluent. The results from the modified model were compared with experimental data, validating the accuracy of the model.

Flow reversal in narrow coolant channels is a critical phenomenon for the safety of research reactors, especially those designed with a downward nominal flow. During a loss of forced flow event, the downward movement of the coolant may temporarily halt before transitioning to an upward natural circulation. Fuel damage can result if dryout occurs and the fuel or cladding temperatures exceed safe limits. This study provides an in-depth examination of flow reversal in narrow rectangular channels by analyzing experimental results and an analytical approach. The analytical approach utilizes the calculated steady-state system pressure drop versus flow curves for different heat flux values. A review of the relevant literature was conducted, and selected experimental data were utilized to benchmark against the flow reversal limits predicted by the analytical approach. The experimental data is selected from tests involving flow reversal in a narrow rectangular channel. The findings were compared with successful flow reversal test data and predicted dryout power under dryout conditions. Also, the study examined the effects of inlet liquid temperature, system pressure, and localized pressure drops. The analytical approach provides physical insights into the flow reversal phenomenon. The approach may be used in conjunction with thermal-hydraulic analysis software, strengthening the confidence in the software predictions.

The present study presents a numerical investigation into the fluid flow and heat transfer performance of a straight mini-channel with an additional inter-connected mixing area in a heat sink plate. The influence of the dimension of the inter-connected area on the thermal-hydraulic performance was examined. Three different sizes of the inter-connected area, defined in terms of aspect ratio (AR), were studied to understand their effect on the thermal-hydraulic performance. Water was used as the coolant, flowing in a single-phase regime under turbulent conditions at Reynolds numbers ranging from 1000 to 4000. A constant heat flux of 10 kW/m2 was applied to both surfaces of the cooling plate. The grid independence test showed a deviation of less than 2% for the friction factor and Nusselt number, while the GCI results indicated that the deviation of the friction factor and Nusselt number was less than 2% within an asymptotic range around 1. The results demonstrated that the aspect ratio of the inter-connected area has an impact on the thermal and hydraulic performance. Both the friction factor and Nusselt number decreased with an increase in the size of the inter-connected area. Furthermore, this study revealed that the interconnection zone created two stationary circulation zones, which influenced the velocity and temperature contours. Finally, a new correlation was developed to explain the relationship between the friction factor and Nusselt number in terms of the Reynolds number and the aspect ratio of the inter-connected area.

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

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

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

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

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

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

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

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

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

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

Subchannel (SC) analysis has historically supported core design numerical simulations for a wide variety of concepts encompassing thermal and fast reactors. The suitability of SC codes to heavy liquid metal coolants and extreme operating conditions have been the object of a work package in the framework of the IAEA Coordinated Research Project ‘Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop (NACIE)-CRP-I31038’, based on the experimental data provided by the NACIE-UP loop located at the ENEA.

The facility features 19, electrically heated, Fuel Pins Simulator (FPS) arranged with a triangular pitch and spaced by a wire wrap, instrumented with 67 thermocouples. The thermal hydraulic problem in the FPS assembly has been simulated by eleven institutions with nine different SC codes. The focus of the simulations is on the steady states of both forced and natural circulation conditions, as well as the in-between transition. In this paper two extreme cases ahave been analysed, one with all pins heated (ADP10) and one with only the seven central pins active (ADP06). ADP10 test is more representative of a condition which could be found in power reactors, while ADP06 test is a challenging power-profile condition, which allows for unprecedented physics insights, both cases can be used for validation of the SC codes .

The comparison demonstrates that SC codes can reliably capture the temperature profile within a wire-wrapped pin assembly, though it also highlights a need for modelling improvements of the wire effect with extreme intra-bundle temperature gradients.

TerraPower is developing a subchannel analysis code, Mongoose++, to support the design of the Natrium^® demonstration reactor. Development of this code was initially motivated by the need to predict core-wide duct temperature distributions to calculate reactivity feedback from radial expansion and assembly bowing. Such calculations require resolution of local coolant flow and temperature distributions within an assembly as well as global inter-assembly heat transfer effects. While Computational Fluid Dynamics (CFD) remains computationally prohibitive for routine design and analysis calculations at the core-wide scale, intermediate fidelity subchannel methods are well-suited for these tasks and have a proven record in the industry for licensing calculations.

Mongoose++ is written in C++ and traces its lineage to the legacy COBRA series of subchannel codes. Mongoose++ utilizes a similar formulation of the subchannel conservation equations but with several advancements to support the specific needs of the Natrium^® project. The subchannel equations are discretized using the finite volume method on a staggered mesh and solved iteratively using a variant of the SIMPLE algorithm. Each iteration of the SIMPLE algorithm is parallelized across assemblies. While more costly than traditional axial-marching schemes, the SIMPLE algorithm is more robust when modeling assemblies with significant buoyancy-induced flow redistribution and localized flow recirculation; such conditions may arise in SFR non-fuel assemblies during off-normal operating conditions at reduced flow.  

This paper provides a detailed overview of the design and implementation of Mongoose++, discussing current capabilities and planned developments. Benchmark comparisons against legacy experimental data and recent CFD calculations are presented.

Recent efforts have focused on at establishing high-fidelity, multi-physics safety analysis methodologies to assess realistic safety margins. However, these approaches still exhibit conservatism by employing conservative initial conditions, such as assuming a hot rod with maximum power for the whole reactor core. This study presents the development of a coupled code for pin-wise reactor core analysis, coupling the subchannel analysis code CUPID, neutron kinetics code MASTER, and fuel performance code GIFT. The objective is to generate accurate pin-wise fuel rod conditions during normal operation, providing more realistic initial conditions for safety analysis. The coupled code performs thermal-hydraulic analysis of the reactor core, accounting for fuel deformation, and simulates realistic power distributions with reactivity feedback from coolant and fuel temperatures. Coupling between the codes was achieved using socket communication and dynamic link library (DLL). A practical simulation was performed on an OPR1000 reactor core during the first cycle, and key results from the steady-state simulation were evaluated. The impact of the fuel performance code was also examined by comparing the results of the coupled CUPID/MASTER and CUPID/MASTER/GIFT codes. Finally, the effect of pin-wise initial conditions on safety analysis was investigated for a reactivity-initiated accident (RIA) scenario with results compared between conservative initial conditions and pin-wise initial conditions.

Hitachi has been developing subchannel analysis codes to predict thermal-hydraulic characteristics of newly designed BWR (Boiling Water Reactor) fuel bundles. The steady-state subchannel analysis code SILFEED (Simulation of Liquid Film Evaporation, Entrainment, and Deposition) with an updated film flow model has been mainly utilized for mechanistic film dryout predictions of various fuel bundle designs. Next-generation fuel bundles, such as Hitachi's RBWR (Resource-Renewable BWR) and GNF3, feature tight lattice configurations, axially varying water rod, and partial-length rods. For these designs with complex geometry, accurate evaluation of thermal hydraulic behavior for each subchannel is required.

To address these demands, Hitachi is developing a new subchannel analysis code based on a transient two-fluid three-field model enabling more advanced evaluations of void fraction and pressure drop by directly solving the void fraction. While its current capabilities are limited to steady-state and transient void fraction and pressure drop, future developments aim to extend its capability to fuel temperature and critical power predictions.

The first step of the code validation, we performed calculations of void fraction and pressure drop in NUPEC 8×8 bundle and φ5.2 to 10.1 mm tube, and compared with experimental data. The results demonstrate that the code achieves prediction accuracy of void fraction within ±15% and pressure drop also within ±15% under BWR operating conditions, which are comparable to those of other subchannel analysis codes.

Accurate modelling of the annular film flow regime in BWR fuel assemblies is of paramount importance for the prediction of the minimum film thickness, dryout location and duration, and crud deposition. Recent decades have seen the introduction of complex assembly structures such as part-length rods and spacer grids, whose complex effects on the flow must also be modelled appropriately for accurate estimation of safety parameters. This paper presents the initial steps towards an improved modelling package for churn-turbulent and annular flow at a subchannel scale. The subchannel code CTF, which uses a two-phase three-field approach, is modified to implement updated models for the droplet entrainment and deposition rates and a new solver (SCARF) has been developed, verified, and applied to partition the subchannel flow rate of the liquid field from CTF during annular flow into distinct films on adjacent rods. This method enables the assessment of film depletion on the sides of each rod and preserves the relative isolation of the films. The results of the two codes are then compared and show that the SCARF produces similar estimates of global flow parameters significant differences at the subchannel scale are observed. The effect is exacerbated for non-uniform radial power profiles. Future research will expand on this by incorporating new models for turbulence and void drift. Additionally, these models will consider how the distribution of film flow within the assembly affects inter-channel and inter-field transfer rates.

Plate-fuel reactors are one of the most common types of research reactors. The thin fuel plates have characteristics such as enhanced thermal conductivity and heat transfer properties, optimized neutron flux, and fuel performance, which makes them an attractive alternative. CTF is a state-of-the-art sub-channel code used for reactor thermal-hydraulics, initially developed for rod bundles and core analysis. In the current version of CTF, the heat generative solid structures are limited to cylindrical shaped rods or tubes. This work aims to develop a new heat structure model in CTF for the plate-type fuel geometry expanding the current capabilities of the code. This new feature extends the applicability of CTF to research reactors, mini- and micro reactors utilizing this fuel design. This development supports the growing demand for accurate thermal-hydraulic modeling and simulation of diverse fuel types and configurations to support the feasibility and safety analysis of small modular reactors. The proposed paper details the implementation process of the plate-fuel heat structure in CTF together with a detailed verification and validation process based on the analytical solution and the experimental available data, respectively. The verification process includes heat conduction within the plates and the convective heat transfer between the plate and the coolant. Thermal expansion and fuel performance models have been analyzed and they are currently suggested as a future improvement of the code.

In the event of a severe nuclear power plant accident, radioactive materials may be released into the environment as aerosols and transported over long distances by atmospheric motion, posing significant risks to human health and the environment. In order to accurately characterize the temporal and spatial dynamics of radioactive aerosol dispersion after a nuclear power plant accident, a dispersion numerical simulation method based on the Euler-Euler model was proposed. An aerosol concentration distribution solver was independently developed based on the open-source computational fluid dynamics (CFD) platform to realize the numerical simulation calculation of aerosol concentration distribution in a large space. The stable release and dispersion process of cesium iodide (CsI) aerosol in different environmental wind fields was studied. The results showed that the aerosol spreads in the wind field near the ground, and the concentration was always high in the area within 400 meters from the source (the concentration was higher than 1.47 × 10¹⁶ m^-3). With 1.47 × 10¹⁶ m^-3 as the detection standard, in a 5m/s ambient wind field, aerosol spreads to 3660 meters downstream of the source at 800s. Assuming that the evacuation time is 10 minutes, the danger degree is highest within 1.77km of the source as the center, and residential areas are not suitable within 4km.

The SCENES program is an integrated software package for nuclear power plant design and safety analysis independently developed by Shanghai Jiao Tong University. In order to verify the accident analysis ability of its system analysis code SCENES-netFlow, this paper selects ACME bench for modeling analysis. The ACME test facility is based on CAP1400 and is mainly used to verify the safety of the passive system in the event of small break LOCA and non-LOCA in the prototype power plant. This paper mainly analyzes the working conditions of CAP03 small break LOCA. The results show that the predicted accident sequence and test phenomena are consistent with the experience. The main results of numerical analysis can well reflect the experimental phenomena and agree well with the experimental results, indicating that SCENES-netFlow has the ability to simulate the accident conditions.

In the event of a severe accident at a nuclear facility, molten core components, known as corium, can form and pose a risk if not properly cooled. Corium can breach the reactor pressure vessel and cause the ablation of containment concrete in a process called Molten Corium Concrete Interaction (MCCI). Understanding MCCI is essential for evaluating containment safety, with previous studies using experimental and numerical approaches. Traditionally, system codes and lumped parameter methods have been employed, while CFD simulations have largely focused on corium spreading. This study introduces a novel approach using a multiphase flow technique to predict natural convection and phase change processes in MCCI. The model integrates multiphase heat transfer, phase change, mass mixing, and decay heat generation, implemented in the OpenFOAM CFD code. It is validated against a PCM melting experiment, showing excellent agreement with experimental data. The validated model is then applied to simulate the COMET-L2 experiment, incorporating decay heat sources and phase changes. A mesh sensitivity study and time-step variations are conducted for model convergence, with results closely matching experimental data. Detailed analysis of concrete ablation, crust formation, oxide relocation, and metal penetration into the basemat is provided, offering insights into the thermal behavior of corium and concrete during MCCI. This approach enhances the understanding of MCCI phenomena and supports improved safety assessments for nuclear containment.

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

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

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

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

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

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

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

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

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

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

The lead-cooled fast reactor (LFR) is one of the promising Gen-IV designs which is being actively developed by several commercial vendors. One of the safety aspects of interest for LFR designs is the potential burst of a fuel pin.

During a fuel pin failure event, a jet of gaseous fission products is ejected into the coolant subchannels adjacent to the failed pin. The jet has a relatively high momentum due to the pre-pressurized nature of fuel pins. The question of interest is whether the gas jet and subsequent gas bubble formation in coolant subchannels could potentially thermally blanket adjacent fuel pins, leading to them failing. Hence, a potential chain failure propagation across the core is imaginable.

The aim of the present work is to experimentally investigate the bubble formation, location and behavior.

The experimental setup used for the investigation consists of a liquid metal loop equipped with high-resolution measurement techniques. First, experiments were conducted using a single sub-channel test section, combined with high-speed imaging. This experimental campaign was used to gain novel first general insights into the behavior of a buoyant gas jet in low Prandtl, high-density liquid. A second test section was then built, allowing for a multichannel observation of the phenomenon. The paper will include the results of the two experimental campaigns and the conclusions that could be drawn concerning the potential occurrence of fuel pin failure propagation in LFR cores.

After the SGTR (steam generator tube rupture) accident in the LFRs, the high-pressure water on the secondary side enters the primary side and is heated to generate a large number of bubbles, which may hinder the flow of LBE in the reactor core, cause heat transfer deterioration, and threaten the nuclear safety. The behavior of bubbles in the fluid phase is obviously affected by the size of the flow channel. However, there have been relatively few systematic studies on the influence of channel size on bubble distribution characteristics in LBE. The drift flux model is one of the most successful models to predict the distribution of void fraction in gas-liquid two-phase flow. In this paper, based on the upward flow experiment of gas-LBE two-phase flow in circular pipes with various hydraulic diameters, the phase distribution parameters were measured, and the influence of channel size on the parameters of drift flow model was studied.

After the occurrence of a steam generator tube rupture (SGTR) accident in lead-bismuth fast reactors, the interface evolution and transport of bubbles can lead to bubble aggregation, thereby affecting core safety. The interfacial area transport equation (IATE) is an important method for predicting IAC and has significant applications in system analysis codes. In this study, a nitrogen-liquid lead-bismuth metal two-phase flow experiment was conducted in a vertical circular tube channel, and the local interfacial area concentration (IAC) was measured. The measurement data reflect the radial distribution and axial development characteristics of IAC and reveal the evolution and transport characteristics of interfaces. Additionally, this study reviewed the available IAC prediction models including IAC correlations and IATE. However, most of these prediction models have been not developed for the gas-liquid metal two-phase flow, the experiment database was used to verify the applicability of these models in the liquid lead-bismuth metal fluid. The verification shows that the IAC correlations cannot give good predictions of the IAC in liquid lead-bismuth two-phase flow, while the IATE could have a better prediction result, but there is still some difference from the experimental measurement IAC.

During a steam generator tube rupture (SGTR) accident in a lead-cooled fast reactor (LFR), vapor entering the core can induce power excursion and threaten reactor safety. Accurately predicting bubble transport in LFR during SGTR is crucial for its safety assessment. This paper uses a neural network (NN) to predict the bubble distribution within the Europe Lead Cooling System primary system during SGTR accidents. The NN-based model uses one-hot encoding to accommodate heterogeneous inputs and implements a modified Softmax function to avoid non-physical outputs. The method of deep ensembles then quantifies the prediction model uncertainties. The prediction model can accurately predict bubble distributions at three different locations. A relatively large ensemble size is required to converge the ensemble mean, while the convergence of ensemble standard deviation may suffer from outlier samples. Ensemble predictions at different locations tended to be negatively correlated, which usually became weak near extreme values (0 and 1).

The steam generator heat transfer tube rupture (SGTR) accident can lead to violent interactions between lead-bismuth liquid metal (LBE) and water in lead-cooled fast reactors, which can seriously threaten the safety of the core. In this paper, high parameter experiments and refined numerical simulations are used to investigate the lead-bismuth liquid metal-water interaction mechanism. Thermocouples and pressure sensors were used to capture the fluctuation behavior in temperature and pressure in the experiments. This complex and opaque internal interaction is modeled by constructing dynamic boundary conditions of multiphase and multiphysics processes. We demonstrated the existence of three stepwise sequential interaction mechanisms. Moreover, special phenomena such as vapor film wrapping around the core of the jet and secondary penetration have been discovered. his study provides new insights into the interaction between LBE and water and offers important reference for developing mitigation strategies for SGTR.

This study investigates the optimization characteristics of a secondary-side passive residual heat removal system (PRHRS) for enhancing inherent safety performance in lead-bismuth cooled fast reactors (LFRs). Using the fully implicit NUSOL-LMR code with fluid-structure coupling, analyses demonstrate that the PRHRS activates secondary-side natural circulation during unprotected transient overpower (UTOP) and unprotected loss of heat sink (ULOHS) accidents, reducing core temperature rise by 12.3% (p<0.01) while maintaining fuel temperatures 231 K below melting thresholds. The system synergizes with inherent reactivity feedback (coolant density feedback: −1.22 pcm/°C, Doppler feedback: −0.663 pcm/°C) to suppress coolant solidification risks. Under UTOP conditions, PRHRS caps peak cladding temperatures at 1,147 K (52% below safety limits), whereas during ULOHS, it sustains decay heat removal via secondary-side passive flow (2.3% rated capacity). Results conclusively show that integrating the passive system significantly enhances inherent safety under extreme accidents, substantially mitigating potential risks. These findings provide critical insights for optimizing safety designs in forced-circulation LFRs.

The demand for nuclear energy is rapidly increasing and, as such, the deployment of new reactors must be accelerated. Microreactors are being designed to provide electricity and heating for remote applications. Operating the microreactors in these applications will take advantage of remote and autonomous systems, which will be based on digital twin simulations. The development and validation of real-time digital twins is therefore a necessary component for microreactor deployment. The intended coupling of the digital twin control system to hardware provides opportunities to leverage methods that combine hardware and software. Hardware-in-the-Loop (HIL) testing can be integrated with a digital twin, in which an experimental subsystem will replace a region of the digital twin. As-built components can be tested directly, reducing time spent in intermediate modeling steps, and can be used for validation by providing real data for a portion of the reactor.

This paper will present the experimental setup, model development, and results of HIL testing combined with a digital twin of a microreactor cooled with heat pipes. The experimental subsystem will provide thermal hydraulic data of a hexagonal unit cell made up of a heat pipe and the surrounding region. The design of the experimental subsystem and the digital twin will be optimized to demonstrate the method and will not represent an existing microreactor. Validation of the digital twin will be performed by replacing subregions of the core with the experimental subsystem in the digital twin and comparing the results with those from the digital twin alone.

Heat pipes are passive two-phase heat transfer devices utilized in applications such as core cooling for nuclear microreactors, high-efficiency heat exchangers, and other advanced energy systems. The two-phase flow and heat transfer dynamics within heat pipes are often highly complex, particularly during transients and under vertical operating conditions. The present work develops a comprehensive heat pipe transient experimental database for a vertical heat pipe of approximately 2 meters in length using water as the working fluid, with the reported data including internal measurements of operating pressures, pressure drops, liquid film temperatures, evaporator wall temperatures, and vapor core temperatures. In particular, vapor core temperatures were obtained using a fiber optic distributed temperature sensor running along the entire heat pipe length. The database includes power input and condenser coolant flow rate transients to enable the evaluation of the heat pipe’s response to changes in both evaporator and condenser conditions. Experiments were conducted for two different wick types: annulus-screen and wrapped-screen. Important phenomena identified include vapor generation in the annulus and the presence of a subcooled liquid plug near the condenser endcap. The data obtained can be readily used for verification and validation of numerical modeling tools under development for heat pipe microreactor analysis.

Heat pipe cooled reactor (HPCR) is in the spotlight as one of the advanced reactor with the advantages of high inherent safety and compactness. HPCR was originally designed for application in space, low-efficiency power conversion systems were applied. Recently, the HPCR system for power generation has been actively studied with high efficiency power conversion system such as supercritical CO2 Brayton cycle. However, due to compact size of HPCR, there was a problem that the size of the intermediate heat exchanger that satisfies the core power increased. In this regard, we suggested new design of compact intermediate heat exchanger based on printed circuit heat exchanger (PCHE). The heat exchanger consisted of “heat pipe layers” in which heat pipes were inserted and “cooling channel layers” in which the cooling channels were machined. The structural integrity of each layer was evaluated based on ASME standard, and flow uniformity was evaluated through CFD. Based on the temperature distributions of the heater for heat pipe simulation and the heat exchanger body which were measure with fiber optic sensor (FOS), thermal contact resistance and overall thermal resistance of heat exchanger were measured. Through this study, the transportability of the designed heat exchanger was evaluated, and the possibility of comprehensive analysis through integration with the heat pipe and the reactor core model was confirmed.

Heat pipe-cooled microreactors (HPMRs) utilize alkali metal heat pipes for efficient and passive heat transfer. Simulating startup and shutdown of HPMRs require accurate modeling of transient heat pipe behavior. In this study, a 2-D transient heat pipe analysis code, SNUHTP, was developed with a melting/solidification model to simulate frozen startup and phase change effects. Transient verification against an analytical lumped model and steady-state validation using sodium heat pipe experiments showed good agreement in normal operation range of sodium heat pipe. The melting/solidification model was verified with a 1-D Stefan problem, and transient validation using the SAFE-30 heat pipe experiment showed delay in temperature rise due to latent heat effects. The results demonstrate that SNUHTP effectively predicts transient and steady-state heat pipe behavior, supporting its application to HPMR analysis.

The heat pipe-cooled reactor offers numerous advantages, including a compact design, high power density, exceptional reliability, and intrinsic safety features, making it a promising candidate for future mobile power generation systems. This reactor employs a solid-core design where high-temperature heat pipes establish a direct link between the reactor core and the energy conversion system, creating a compact and modular configuration. Despite its advantages, the intricate multiphysics interactions within the system pose considerable challenges for comprehensive analysis. To tackle this issue, this study proposes a multiphysics coupling analysis framework tailored for heat pipe-cooled reactors, developed within the OpenFOAM platform. The framework integrates neutron physics, core thermal transfer, heat pipe dynamics, and thermoelectric conversion models. Its accuracy is verified against experimental data from the KRUSTY space heat pipe reactor's ground-based nuclear testing. A complete system simulation of KRUSTY is then conducted, emphasizing the interplay of multiphysics phenomena under nuclear thermal power conditions.

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.

Small Modular Reactors (SMRs) rely on passive safety systems, which use gravity-driven natural circulation to transfer heat from the core to a large reservoir without operator intervention for extended periods of time. Some SMR designs can reach Rayleigh numbers around 10^15, but the limited experimental data at this scale highlights the need for further validation of predictive thermal hydraulic codes.

This study presents preliminary Computational Fluid Dynamics (CFD) calculations to support the design of an experimental rig at the PANDA facility at PSI, Switzerland, under the OECD/NEA PANDA project. Using a 2D axisymmetric CFD model created with EDF's open-source Code_Saturne software, parametric studies were conducted to investigate the heat transfer mechanisms. Key variables, such as geometrical configuration and dimensions, operating conditions, and numerical options, were examined. Initial results indicate that achieving a Rayleigh number of 10^15 is feasible under the constraints of the facility, with favorable alignment to existing Nusselt-Rayleigh experimental correlations at lower Rayleigh numbers. The computational results allowed the sizing of the components and fixing the operating conditions. These results also underscore the importance of a detailed temperature measurement strategy, particularly near the wall, to be used for the validation and verification of the models.

Further studies explore the influence of different turbulence models on the results, revealing notable differences in predicting the laminar-to-turbulent transition zone and the peak wall temperature. These findings are useful for ensuring the validity of Nusselt-Rayleigh correlations over a wider range of applicability and allowing for accurate modeling of large-scale passive safety systems.

The conventional containment vessel of nuclear power plant acts as a barrier against radiological impacts and missile threats, independent of the reactor’s cooling. However, as the size becomes more compact, the containment vessel of an SMR roles as the final passive cooling system. The cooling capability around the dome must be ensured during severe accidents, such as Loss of Coolant Accident, to minimize coolant loss through condensation on the dome. Most condensation occurs in the upper dome of the containment vessel, however, studies on the external cooling capability of the dome reflecting the large Ra of SMR have not been conducted. Therefore, this study aims to analyze the heat transfer of domes under high Ra conditions, considering their geometric characteristics and flow behavior. Experimental range corresponds to 10⁹≤Ra_Db≤10¹³. The base diameter of dome, corresponding to the diameter of the cylindrical lower structure, is fixed, but its height can vary. These characteristics of the dome can be defined by a truncation angle, as the dome represents a segment of a sphere. The smaller the truncation angle, the closer the dome resembles a flat plate, while a truncation angle of 90° represents a hemispherical shape. Depending on the truncation angle, the slope of the dome surface varies, leading to differences in flow behavior. Plume flow enhances heat transfer and the separation point, where the point plume flow develops, is varied by the surface slope. Smaller truncation angles result in an earlier development of the separation point and further enhance heat transfer.

Due to the nonlinearity of the Natural Circulation (NC) process, flow and temperature perturbations in an NC loop result in a feedback mechanism. Under certain conditions, the prevailing circulatory flow can become unstable and ultimately stall, leading to insufficient heat transfer. This scenario is of particular interest to PWRs, since inadequate passive heat removal could result in initiation of postulated accident scenarios. Therefore, accurate predictive capability for stall in perturbed NC scenarios is of great interest to this work.

Performing full-scale high-fidelity simulations of the plant during perturbed scenarios is computationally prohibitive from a design basis analysis perspective. However, the phenomena of interest during perturbed scenarios are highly complex and multiscale; current 1D System Codes are not capable of adequately capturing the prevailing 3D flow phenomena. The current work considers a more computationally feasible approach. We explore surrogate modelling via Gaussian Process Regression as a method for accurately predicting bulk flow and therefore NC stall. The surrogate model is trained by considering Unsteady Reynolds Averaged Navier-Stokes (URANS) Computational Fluid Dynamics (CFD) simulations of NC in a simple experimental loop geometry containing an n-bend and multiple heat sinks. Various perturbed NC simulations are presented, in which we consider multiple transient definitions which are specified by an injection flow rate profile at a specified location, heater input power, and n-bend height. We then assess the surrogate modelling approach’s ability to capture the key features of quantities of interest by benchmarking the results against equivalent URANS CFD simulations.

As the primary mechanism of passive heat removal in nuclear reactors, understanding natural circulation is paramount to their safe operation and shutdown conditions. Natural circulation in molten salt reactors is of particular relevance due to the fluids high Prandtl number as well as the phase of the fissile material in liquid-fueled reactors. Natural circulation loops are employed to study the thermal hydraulic behavior of fluids when subject to thermal gradients and small flow disturbances. The objective of this work is to introduce and analyze the effects of local heating conditions on the velocity profile and near wall behavior (such as boundary layer thickness) of molten salts in a heated transparent test section. Particle image velocimetry (PIV) was performed, and the boundary layer was analyzed for three different heating conditions. Those conditions were applied to the transparent test section: a cooling test condition, a thermally isolated test condition, and a heated test condition. In addition, the other thermal conditions of the loop were held constant. As the test section is heated, the peak velocity and slope of the velocity profile increase with test section heater power. Additionally, preliminary transient thermal analyses are presented in this work.

The passive molten salt fast reactor (PMFR), under development in Korea, employs a helium bubbling system to remove insoluble fission products (IFPs). Notably, the helium injection significantly changes entire fluidic performance within the primary system. These changes, in turn, influence heat transfer efficiency. Accordingly, a lab-scale experiment needs to be performed to evaluate the impact of helium injection under PMFR conditions.

To this end, a two-phase natural circulation molten salt loop at a reduced scale was designed based on scaling laws to simulate the helium bubbling effect in the PMFR. New governing equations for the prototype (PMFR) and reduced model (molten salt loop) were established based on the drift-flux model. Similarity criteria were derived through the nondimensionalization of the new governing equations. These criteria were employed in the design of the reduced model. Furthermore, an enlarged model, whose geometrical shape is similar to the reduced model, was also designed to evaluate the effect of friction number.

Subsequently, the distortion among prototype and two models was evaluated using multiphaseEulerFOAM solver in OpenFOAM. The distortion analysis revealed significant discrepancies in void fraction and fluid velocity between the prototype and the reduced model. A major reason for the distortion is attributed to geometrical differences. However, the distortion between the reduced model and the enlarged model was relatively minor due to geometric similarity and friction number scaling. This study will contribute to developing advanced scaling laws applicable to molten salt reactors (MSRs), although there are still rooms for improvement.

The steady-state behaviour of two-phase natural circulation systems (TPNCs) is critical for their efficient and safe operation across various applications, including nuclear reactors, thermal power plants, and advanced passive cooling systems. In TPNCs, steady-state conditions represent a balance between the buoyancy driving force and the opposing frictional force. Understanding and predicting the steady-state characteristics are vital for optimizing system performance, particularly in Boiling Water Reactors (BWRs) and natural circulation boilers (NCBs), where high flow rates are desirable for enhancing the heat transport capability.

This paper provides an in-depth analysis of the steady-state behaviour in TPNCs, focusing on key factors such as loop flow regimes, the effects of system geometry and heat input on circulation patterns. The study examines the influence of system parameters like loop diameter, heat flux and pressure on steady state flow. In addition, the effect of loop inventory on the steady state performance has been studied. The predictions cover the inventory at which peak TPNC flow rate, breakdown of TPNC flow and heat-up of the heated surface occurs. Insights gained from this analysis are crucial for designing TPNC systems to maximize the heat transport capability for critical applications such as nuclear reactors and thermal power plants.

A new experimental setup has been constructed for a cold swirling turbulent jet issued through a pressure-swirl atomizer generating a spray at Reynolds number up to Re=10⁶.

The cold spray injection is used for steam condensation and pressure regulation in the pressurizer of a pressurized water reactor (PWR).

For large spray flowrates the droplets also reach the pressurizer tank wall, which acts as an undesired thermal load.

Current simplified prediction tools for transient load calculations lead to conservative estimations of the loads and under predicted lifetime.

More advanced tools, e.g. computational fluid dynamics (CFD) require better models for two-phase flow heat transfer in order to get more reliable lifetime predictions in a long term operation (LTO) context.

Measurements have been conducted in a 1:1.84 lab scale model of the spray two-phase flow characterizing the liquid fraction, droplet size and velocity distributions dependence on the spray flow rate and surface tension.

The spray cooling heat transfer has also been measured using a unique heat transfer sensor developed at CEA in France.

The experimental data base will be used for validation of more advance CFD models, being developed in conjunction to the present experimental campaign.

With the development of surface engineering, porous surfaces have emerged as a significant research subject in boiling heat transfer. The latter, in turn, plays a crucial role in various industries such as power plants, distillation plants, and microelectronic technology. In this paper, the Molecular Dynamics method is adopted to investigate the wicking dynamics and boiling dynamics of two porous surfaces: foam, which exhibits randomly distributed pores, and mesh, composed of ordered square wires with relatively uniform pore sizes. Three wettability, namely hydrophilic, neutral, and hydrophobic wetting states, are assigned to the two porous surfaces de-coupling the effect of wettability from surface structure. Results reveal that, during the wicking process, the foam surface shows better wetting ability as it absorbs liquid under both hydrophilic and neutral wettability. Comparatively, the mesh surface has the fastest wicking speed under hydrophilic wettability yet it becomes non-wetting under neutral wettability. During the boiling process, the boiling dynamics differ greatly under three wettability. More importantly, the difference in surface structure makes the foam surface possess a better heat transfer whereas the mesh surface causes gentle pressure variation. Our findings provide insights into the design of artificial porous surfaces for certain purpose and their potential application.

Water entrainment at T-junctions is of upmost importance in some nuclear safety analyses. Such a phenomenon directly impacts the core liquid inventory, hence its coolability. Numerical modelling using system-scale codes is essential for characterizing the two-phase flow interactions at the T-junction and in the branch line upstream. In light of this, code validation must be carried out through comparison to experimental data (Separate Effect Tests).

The test section represents, at a lower scale compared to reactor scale, an upper core plenum, a hot leg, and a pressuriser surge line. The test loop is operated at atmospheric conditions. Thus, the transposition issue (geometry and thermal hydraulics conditions) also has to be tackled.

The aim of this paper is to present the set of calculations, the comparison to experimental data and the scaling approach through the confrontation of CFD and system-scale code predictions.

The system-scale computations are performed with the thermal hydraulics code CATHARE and the CFD calculations with NEPTUNE_CFD, an in-house code.

CATHARE and NEPTUNE_CFD results are first compared to the experimental data, both, qualitatively (video recording) and quantitatively (water height) for two configurations (vertical upward and inclined T-junctions). This allows an assessment of the codes’ accuracy regarding the phenomenon of water entrainment at a T-junction and raises reflections on the physics, and the modelling.

Then, CATHARE and NEPTUNE_CFD calculations are performed at reactor scale and thermal hydraulics conditions. It is assumed that CFD better copes with scaling and is taken as a reference for the code-to-code comparison.

Finally, future work is proposed.

The efficient heat dissipation of the permanent magnet propulsion motor rotor is crucial to the development of advanced propulsion systems. As an advanced thermal management technology, the rotating heat pipe enables effective cooling of rotating components through internal phase-change heat transfer and natural circulation. Based on this research background, our team has built a rotating heat pipe experimental system and conducted experiments with a 70% filling ratio, a length of 500mm, and a diameter of 30mm using a stepped rotating heat pipe. The results show that its heat transfer capability gradually increases with the rotational speed. Under the same conditions, the heat transfer performance of the parallel-axis rotating heat pipe reached twice that of the coaxial rotating heat pipe, with an equivalent thermal conductivity of up to 1438.08 W/(m·K). This study provides experimental data support for the application of rotating heat pipes in the cooling of permanent magnet propulsion motor rotors.

The pool boiling experiment for the observation of corrosion products deposition is carried out to better understand the fouling mechanism under nucleate boiling conditions. The experimental apparatus comprises the quartz glass cavity, test piece (aluminum), high-speed camera and heater set-up. The deposition tests are performed in dilute colloidal solution (Fe₃O₄) with different wall temperature and bulk temperature at atmospheric pressure. The experimental observations indicate that the deposits exhibit a circular distribution and a thickness of approximately a few micrometers under nucleate boiling. The fouling ring is distinguished by a lower central thickness and a higher edge thickness. To gain further insight into the flow field distribution during the bubble growth process, the numerical simulation of the bubble growth and detachment process is conducted using the CFD method. It has been demonstrated that corrosion products are transported to the contact line of the bubble as a consequence of turbulence vortex. Besides, the micro-layer situated at the base of the bubble will undergo a process from thinning to drying out, resulting in the deposition of corrosion products on the heated surface. Through a combination of experimental and numerical techniques, the transport mechanism of corrosion products under the influence of nucleate bubbles has been elucidated, and the model between the evaporation flux and deposition rate of corrosion products has been developed.

The Best Estimate Plus Uncertainty (BEPU) methodology, developed over several decades, has seen numerous innovations aimed at enhancing the efficiency and quality of the BEPU procedure. Wilks’ formula characterized by nonparametric statistics is widely used for uncertainty evaluation, while it is time consuming. Deterministic sampling (DS) methodology assesses the uncertainty of outputs through the first two orders of moments of the input uncertain parameters. The reduction in computational effort achieved by using fewer sampling times, compared to the Wilks method, presents a promising alternative for enhancing the BEPU methodology. 16 input parameters and 3 safety-related output parameters as the Figure of Merits (FoMs) are chosen in ESBWR initiated by main steamline break for BEPU evaluation using RELAP5. First order Wilks’ method and three DS (DS-Standard, DS-Simplex, and DS-Hadamard) methods are applied. Subsequently, a preliminary sensitivity analysis of the Wilks’ results is performed to identify the input parameters with a significant impact on the FoMs. The downscaled parameters were then used as inputs for BEPU calculations using three DS methods. The degree of envelopment and conservatism of the three results (Wilks’ results with 16 input parameters, three DS results with 16 input parameters, and three DS results with downscaled parameters) relative to the experimental data were compared to determine whether the downscaled input results could be considered valid for the final BEPU analysis under the given conditions.

Steam Generator Tube Rupture (SGTR) is a critical safety event in nuclear power plants, particularly in Pressurized Water Reactors (PWRs). An SGTR occurs when one or more tubes within the steam generator fail, allowing radioactive coolant from the primary circuit to leak into the secondary side, potentially contaminating the secondary steam and elevating radiation levels outside the reactor containment. This study investigates the thermal-hydraulic response of the PKL facility under a Multi-SGTR (MSGTR) scenario. Conducted within the framework of the OECD/NEA ETHARINUS project, Test J5.1 aims to evaluate the system's performance during an MSGTR event. Two experimental runs were executed: in the first, three out of four steam generators (SGs) were assumed to have ruptured, while the second run assumed failure in all four SGs at the PKL facility. The study presents the outcomes of blind simulations for both scenarios, emphasizing the differences in operational sequences due to varying depressurization and cooldown strategies. In Run 1, depressurization was initiated via the intact SG, followed by activation of the pressurizer (PZR) relief valve on the primary side. In Run 2, only the PZR relief valve was used for depressurization. Both runs extended to 30,000 seconds, during which primary and secondary pressures equalized.

This paper presents an experimental investigation of thermal radiation effects in PANDA facility (PSI, Switzerland), focusing on gas mixture atmospheres with different steam content. Thermal radiation influences the containment atmosphere temperature and buoyancy and, therefore, has an impact on the hydrogen distribution during a postulated accident. The experiments were conducted as part of the ongoing efforts to understand the role of radiative heat transfer in containment thermal hydraulics. In this paper, we analyze the experimental results of two new tests of the so-called P1A2 series performed within the OECD/NEA PANDA project.

The test atmosphere initially consisted of an air and steam mixture at 110°C, with steam concentrations ranging from nearly zero to high values (60%). A stratification layer of 50% nominal helium was created in the upper 2 meters to isolate with best efforts the effects of radiative heat transfer from convective mixing. For the compression of the gas mixture, helium was injected at 10 g/s from the top over a period of 1200 seconds. The subsequent gas thermal behavior and concentration distribution were recorded during the compression phase and an 1800-second decay phase. The results demonstrate that the magnitude of the temperatures is a strong function of the initial steam content, with higher temperatures for lower steam content. Additionally, the experiments confirmed that thermal radiation has a major impact on temperature homogenization during the decay phase, with faster homogenization occurring in atmospheres with higher steam content. These findings have profound implications for CFD calculations in the presence of steam.

The direct vessel injection technology is gradually adopted in new pressurized water reactors because of its advantages of simplifying the design of safety injection system and improving economic benefits. This experiment focuses on the modified DVI (direct vessel injection) safety injection system of the HPR1000. An improved linear scaling method was employed to model the prototype, and relevant experiments were conducted on a 1:8.5 scale visualization test section. Through experimentation, phenomena such as CCFL (counter-current flow limitation) and bypass flow near the break were observed during the refilling and reflooding stages within the annular cavity under a large break loss-of-coolant accident (LBLOCA). The study investigated the impact of various break locations, injection heights, and the presence or absence of guiding structures on the bypass effect in the DVI safety injection system. Additionally, comparisons were made between DVI safety injection and cold-leg injection. The research findings reveal that direct bypass flow dominates the safety injection bypass during the refilling and reflooding stages of an LBLOCA. The closer the cold-leg break location is to the DVI nozzle, the significant increase in bypass flow at the break location. Different DVI injection heights affect the spread of the liquid film, thereby influencing the proportion of bypass flow. The installation of guiding devices can effectively reduce the proportion of safety injection bypass flow. The data results from this study provide crucial insights for the optimization and innovation of the modified safety injection system in HPR1000.

In marine environments, the internal components of in-vessel pressurizers, such as control rod guide tubes and electric heating rods, are susceptible to structural damage due to fluid slamming, especially under severe ocean conditions. Therefore, it is crucial to study the structural reliability of the pressurizer.In order to analyze this issue,this study employs a six-degree-of-freedom mobile platform to input various motion excitations. Strain gauges were attached to different locations of the internal components to measure strain responses,and this study conducts a comprehensive time-frequency domain analysis of the structural responses caused by fluid slamming under various operating conditions using the Fast Fourier Transform (FFT) and Continuous Wavelet Transform (CWT). The research results show that the strain on the first layer of the control rod guide tubes in the direction of motion is significantly greater than that at other locations, but it remains below the material's yield limit, and it has little influence on the equipment. In coupled motions, the location of maximum strain is determined by the motion with the largest amplitude, and the strain on internal components located in directions with smaller motion amplitudes is influenced simultaneously by this motion and the motion with the largest amplitude.This study provides insights into the structural response of pressurizer internal components under fluid slamming in marine environments.

To reduce computational expenditures, thermal radiation has often been neglected in system or CFD code simulations of containment flows involving infrared-absorbing water vapor. However, large-scale experiments at PANDA facility, along with related benchmarks using CFD tools have demonstrated that radiative heat transfer is significant, even at very low steam content. Accordingly, the present work focuses on exploring the effects of thermal radiation within a smaller containment environment.

Two tests with different steam volume concentrations (0.1 and 2.5%) were conducted in the medium-scale LINX facility, which is a vessel of 2-meter diameter and 4-meter height (1/10 PANDA drywell volume). Steam/Air mixture was compressed by injecting Helium from the top at a mass flow rate of 1g/s for a duration of 1200 seconds. This led to the formation of a helium layer at the top, which pushed down the steam/air mixture, creating a high-temperature bubble. Additionally, CFD simulations of both tests were performed using ANSYS Fluent, employing the k-ω SST turbulence model and the P1 model to incorporate the thermal radiation. The steam absorptivity was treated with the Weighted Sum of Gray Gases Model (WSGGM).

The experiments showed that 0.1% steam test yielded a significantly higher peak temperature compared to the 2.5% steam test. CFD simulations without inclusion of thermal radiation highly overestimated the temperature profiles. Meanwhile, a good match with experimental data was achieved using the P1 model. Overall, these results highlight the importance of considering thermal radiation when modeling naturally driven flows in steam-containing atmospheres, even at smaller scales.

Understanding the coolant flow behaviour in the primary heat transport system (PHTS) is crucial for reactor safety analysis of accident scenarios. Databases from rigorously designed experiments are necessary to support the modeling of complex coolant flow behaviours. This study focused on analyzing the two-phase flow distribution in an important component of a typical CANDU-type PHTS, namely the header-feeder system, under flow conditions relevant to large break loss-of-coolant accidents (LB-LOCA). Experiments were conducted on a 1:3 scaled Header Test Facility that replicates the piping configuration representative of a CANDU PHTS (specifically the Advanced CANDU Reactor, ACR-1000) header/feeder system using air-water as the working fluids to simulate steam-water. The experiments simulated a scenario where a large break occurs at the inlet header and the emergency core cooling system fails to activate. Flow conditions were varied using a range of water and air flow rates, simulating different levels of a LOCA. A total of 116 wire mesh sensors were installed along the header and feeder pipes, measuring various two-phase flow parameters, including instantaneous void fraction, bubble size and interfacial velocity distribution in the flow channel. Experimental data revealed that the void fraction and air flow rate in each feeder depend not only on the initial break condition and local phenomena in the headers, but also on the location where the feeder is connected to the header and the elevation of the feeder pipe.

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

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

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

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

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

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

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

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

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

The high-temperature gas-cooled reactor (HTGR), combined with a helium turbine cycle, represents a cutting-edge application of advanced nuclear energy technology. A critical component in this system is the printed-circuit heat exchanger (PCHE), which features microchannels (1–2 mm wide) for efficient heat transfer. However, these systems face challenges, particularly from dust particles generated within the reactor. These particles tend to deposit on PCHE surfaces due to the microchannels' narrow dimensions and frequent turns, potentially degrading heat transfer performance and blocking the channels. Understanding the long-term effects of particle deposition on PCHE performance is essential for the sustainable operation of HTGR systems. This study investigates this issue using a near-wall drag model to incorporate shear flow effects and an EA rebound model to simulate particle deposition behavior. A dynamic mesh method was employed to track the evolution of deposition morphology over time. Key findings reveal a non-linear relationship between particle size and deposition fraction, with deposition increasing and then decreasing as particle size grows. Smaller particles deposit primarily at upstream bends, where flow dynamics promote adherence, while larger particles settle downstream after sufficient energy dissipation. Additionally, upstream deposition significantly reduces downstream particle accumulation, influencing deposition distribution patterns.

The shell-and-tube heat exchanger (STHE) is crucial for industrial safety and efficiency. Refined simulations of its operational characteristics often use three-dimensional (3D) and one-dimensional (1D) models to calculate outlet temperatures.3D models are complex and time-consuming meanwhile 1D models are simple but less accurate, and data-driven models lack interpretability and have limited application. To overcome these limitations, this paper proposes a hybrid "mechanism-data" driven model for STHE.Firstly, based on heat transfer and fluid mechanics, a 1D simulation model is built using MWORKS/Modelica, analyzing outlet temperature and pressure changes under steady, transient, and fault conditions. Secondly, operational data from STHE test benches, including flow rates, temperatures, pressures, and outlet temperature simulations from mechanism-driven models, are collected to build a data-driven model using deep learning algorithms. This model captures nonlinear relationships and dynamic characteristics, addressing the mechanism model's inability to observe and describe factors like corrosion and baffles.Combining both models, a hybrid "mechanism-data" driven model is established, offering physical interpretability and high accuracy. By simulating test bench operations in real-time, it detects potential faults, identifying their types and severity, supporting maintenance and management.Experimental validation shows the hybrid model outperforms single models in simulation accuracy and fault diagnosis, accurately reflecting STHE operational status and fault characteristics. Future work will optimize and enhance this model for broader applicability and accuracy across different conditions and STHE types.

The supercritical carbon dioxide (sCO₂) Brayton cycle, as a core component in the design of next-generation nuclear energy systems, emphasizes safety, operational efficiency, and non-proliferation characteristics. Within this framework, the Printed Circuit Heat Exchanger (PCHE) plays a key role in optimizing the heat transfer process under high temperature and high pressure conditions. This paper proposes a numerical design framework for the PCHE, focusing on the reduction of flow non-uniformity through a combination of secondary heads and porous baffles. Computational Fluid Dynamics (CFD) methods are employed to simulate the thermal-hydraulic performance of the heat exchanger, assessing the impact of different geometric parameters on flow distribution and heat transfer efficiency. The results demonstrate that the combination of secondary heads and optimized porous baffles significantly improves flow uniformity, thereby enhancing heat transfer efficiency and reducing pressure drop. This study provides valuable insights for optimizing the thermal-hydraulic performance of heat exchangers in supercritical CO₂ Brayton cycles.

Surrogate-based optimization (SBO) is a powerful approach for the design of the airfoil fin printed circuit heat exchanger (PCHE), which maximizes heat transfer and simultaneously minimizes pressure loss. Existing optimization studies of the airfoil fin PCHE commonly focus on the channel configuration and the fin arrangement. However, a meticulous optimization of the airfoil PCHE may consider the parameterization of the shape of the airfoil fin, while existing optimization methods are insufficient in such cases. To address this issue, the free form deformation (FFD) method is applied to parameterize the airfoil fin shapes and design variables are extracted to control the deformation of the shapes. The airfoil fins are divided into two groups according to the upstream and downstream of the channel. Fins within the same group deform synchronously. The heat transfer rate and pressure drop are employed as the objective functions of the optimization. To improve the optimization efficiency, the Kriging surrogate model is adopted to approximate the relations between the design variables and objective functions. Then, a multi-objective optimization using Non-dominated Sorting Genetic Algorithm-II (NSGA-II) is conducted and the Pareto solutions are obtained. Comprehensive optimal designs are selected on the Pareto front, and the thermal and hydraulic characteristics of the optimized designs have the advantage over those of the PCHE with original airfoil fin shape.

The water-cooled SMR is generally designed for flexible operation. In the performance analysis of nuclear power plants, system codes are used to evaluate plant performance under various Performance-related Design Basis Event (PRDBE) conditions. The load-following operation is one of the major PRDBE conditions. Many water-cooled SMRs employ a helical type once-through steam generator (SG), which produces superheated steam on the secondary side. Unlike conventional steam generators, the helical SG has no concept of water level, making superheated steam pressure a key control parameter in balancing plant pressure during load-following operation.

The helical SG comprises thousands of helical coils surrounding the riser, each with a different helical geometry. These geometric differences affect the heat transfer characteristics, potentially altering the outlet conditions. However, modeling every tube’s unique geometry would be inefficient. Instead, the SG is typically divided into multiple units, and system code calculations are performed on this segmented model.

This study explores how the fineness of the nodalization, or the number of divisions of the SG, affects performance analysis results. Using MARS-KS code, the reference helical steam generator is divided into 5, 10, 15, and 20 units, and steady-state calculations are performed for each case. The focus is on comparing the steam condition at the secondary outlet. It is expected that the outlet conditions of superheated steam are similar across different nodalizations, suggesting that coarse nodalization does not significantly impact analysis results. This is expected to allow for more efficient calculations in large-scale performance scenarios.

The Water-Cooled Lithium-Lead Test Blanket Module (WCLL-TBM) is an essential component in ITER that will provide crucial information for the development of the DEMO driver blanket. Our research aims to build a multi-scale system code for the thermal-hydraulic analysis of the WCLL-TBM. The OpenModelica software is used to develop a robust and modular object-oriented library for the components of the WCLL-TBM and the Water Cooling System (WCS) in this work. The various objects contain modelled thermal flow loops with 0D/1D interconnected components such as pipes, heat ports, orifices and valves. The objects describe the different multi-scales and can be nested and combined to form new objects. Such objects include the Double-Walled Tubes (DWTs), First Wall (FW), Breeding Units (BU), Breeding Module (BM), and larger outer circuits - all of which are designed to have replaceable modules with different levels of fidelity. The code aims at fast and reliable thermal-hydraulic predictions of the WCLL-TBM components and WCS during nominal operating conditions (gauge pressure 15.5 [MPa], inlet temperature 295 [textdegree C], outlet temperature 328 [textdegree C]), as well as the transient response of the system to off normal scenarios by varying certain parameters and loading conditions.

This research utilizes the system-level modeling language Modelica and its open-source libraries, Transform and Hybrid, to develop a natural circulation model for the primary loop of a Low-Temperature Nuclear Heating Reactor(NHR200-II). This model includes components such as the reactor core, coolant channels, heat exchangers , control system model and so on. Based on these models, the steady-state behavior of the primary loop under 100% nominal reactor power conditions was simulated. Also, transient simulation analyses were performed for step and ramp changes at 90% nominal power. The simulation results, when compared with RELAP5 data, demonstrated excellent agreement, confirming the validity and accuracy of using Modelica for simulation modeling. Furthermore, the primary control system model established in this study can regulate the core outlet temperature by controlling the reactivity of the core, and the results show that the reactivity control scheme is feasible. The research work in this paper lays a foundation for using Modelica language to carry out nuclear energy system simulation application.

Different types of medical isotopes are needed for kinds of procedures where many options for production are possible. One particular isotope, Astatine-211 or At-211, can be produced using cyclotron based irradiation where a Bismuth target is converted to At-211. During irradiation, significant heat is generated within the target and appropriate cooling is needed to prevent target melting and increase isotope yield. In support of higher At-211, University of Washington (UW), Oak Ridge National Laboratory (ORNL) and Virginia Commonwealth University (VCU) are collaboratively developing better At-211 target and target holder designs to enable higher isotope production yields. In this study, VCU is focused on the thermal-hydraulics modeling of different target designs including material and geometric parameters to enable UW and ORNL collaborators to hit desired yields. Initially VCU is focused on 0-D modeling using lumped parameter analysis approaches to enable a design space to be developed using multi-objective optimization. This enables the ability to explore both differential holder materials (e.g. aluminum or stainless steel varieties) and coolant channel geometries rapidly to reduce the total number of high-fidelity CFD simulations and experiments. The 0-D model was created in Python to include two energy balance ordinary differential equations (ODEs) to predict the Bismuth and sample holder temperatures during irradiation. The heat generation within the Bismuth target and the coolant conditions were acquired from the UW team and implemented in the model. Using the 0-D model,, we’ve identified several potential improvements to both the sample holder design and coolant channels for follow-on CFD and experimental studies.

Fluid-structure interactions play a critical role in numerous engineering applications, such as jet flows around fuel rods in nuclear reactors. Under specific flow conditions, these interactions can give rise to vortex-induced vibrations (VIV), a phenomenon where large-amplitude oscillations occur due to vortex shedding. VIV poses a significant threat to system stability and can lead to operational failure. Therefore, understanding and controlling VIV is essential to mitigate its detrimental effects.

This study explores the passive control of VIV in a circular cylinder that oscillates freely, using a non-linear energy sink (NES). The NES is designed as a secondary system incorporating linear damping and a key non-linear cubic stiffness component. Simulations are conducted using the Reynolds-averaged Navier–Stokes (RANS) turbulence model with strongly coupled fluid-structure interaction model, utilizing the dynamic response of both the cylinder and the NES, as well as the surrounding fluid flow. By systematically adjusting the sink parameters—mass, spring, and damping—this research investigates their influence on the behavior of the coupled system. The system's response is analyzed at reduced velocity within the lock-in range, where the cylinder's motion synchronizes with vortex shedding. The model is validated against existing data from literature which indicate the optimum values for these parameters to achieve the best performance. Key results are presented in terms of vibration amplitude, drag and lift coefficients, Strouhal number analysis, and vortex visualization, providing insight into the effectiveness of the NES in controlling VIV.

Solid-solid contact thermal conduction is a basic heat transfer problem in thermal power engineering, which is significant for thermal design and safe operation of system equipment, such as high temperature thermal protection of aircrafts, efficient thermal management of space orbits, and superior heat transfer chain of nuclear engineering. Objective to the solid-solid contact thermal conductivity of typical structure based on tubes and holes in heat-pipe nuclear reactor systems, theoretical models of thermal conductivity, mechanics and thermodynamic coupling at the microscopic contact interface were established in the paper, obtaining the interface contact thermal conductivity characteristics under the filling of helium in micro gaps, as well as the influence on the contact thermal resistance for interface temperature and external loads. According to the CFD simulation results under different interface temperature and external loads, the Levenberg-Marquardt algorithm was used to fit a high temperature contact thermal resistance correlation on the cylindrical interface. By selecting appropriate fitting parameters, the R-squared corresponding to the fitting results was greater than 0.95, indicating that the calculation model had a good predictive ability for the contact thermal resistance. It was applicable for the rapid evaluation of contact thermal resistances for the solid-solid interface in engineering design and heat transfer analysis of heat pipe reactors.

The fuel irradiation device serves as a critical platform for conducting nuclear fuel irradiation experiments in research reactors. Its structural design, thermal characteristics, and the arrangement of measurement points at the outlet significantly influence experimental results, thereby affecting the thermal power determination of the device and the evaluation of fuel performance. This study focuses on the HFETR irradiation device, employing a CFD-based three-dimensional high-resolution modeling method to investigate the impacts of outlet sensor placement and structural thermo-mechanical properties on power measurement accuracy. Computational results demonstrate that positioning temperature sensors 130 mm upstream of the physical outlet plane effectively represents the outlet temperature field. From a thermal-hydraulic perspective, an annular gap thickness of 1 mm achieves a coolant flow partitioning of 22% through the bypass channel, with parasitic heat losses limited to 4.7% of the total generated power. This configuration ensures adequate cooling of the samples while avoiding excessive heat leakage.

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

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

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

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

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

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

During a severe accident in a light water nuclear reactor, large amounts of hydrogen could be generated and released into the containment during reactor core degradation. Additional burnable gases (H₂ and CO) may be released into the containment in case of molten corium/concrete interaction. As observed during the Fukushima accidents, H₂ and CO combustion could cause high pressure peaks that could challenge the reactor containments.To prevent this risk, most of the mitigation strategies adopted in European countries are based on the implementation of Passive Autocatalytic Recombiners (PARs). Nevertheless, studies indicate that, despite the installation of PARs, it is difficult to prevent, at all times, the formation of a combustible mixture potentially leading to local flame acceleration. To better understand the phenomena associated with the combustion hazard and to address the issues highlighted after the Fukushima events, such as the explosion hazard inside the venting systems. The AMHYCO project aims to propose innovative enhancements in the way combustible gases are managed in case of a severe accident in operating reactors.

As first step, a critical review of the available literature had been performed with the objective to form the basis for the project regarding (1) PAR efficiency under ex-vessel conditions, (2) existing PWR Emergency Operating Procedures (EOPs) and SAMGs regarding containment risk management (3) H2/CO combustion and the available engineering correlations for combustion risk estimation, (4) equipment and instrumentation surveillance under severe accident conditions. This paper provides a survey on the available literature related to the four topics mentioned above.

Most countries with nuclear power plants have implemented measurement systems to assess hydrogen concentration by extracting air from the containment building during severe accidents. This method samples the atmosphere to determine hydrogen concentration, rather than depending on sensors within the containment, to preserve sensor integrity and ensure accurate readings. However, uncertainties may emerge. Firstly, steam condensation during sampling can change the gas composition ratio. Secondly, the recorded time for hydrogen concentration includes a delay from sampling, which can be compared to the time taken for direct pressure and temperature measurements inside the containment. This time lag may influence flammability predictions based on thermal-hydraulic conditions.
We are evaluating the uncertainties associated with hydrogen concentration measurements. The sampling-type measurement system consists of an accident environmental control module, a hydrogen concentration sensing module, and a data processing module. The control module extracts mixed gases from an experimental setup that simulates an accident environment, filters them through a water tank, and returns them to the experimental vessel. The pool scrubbing method eliminates steam and aerosols. Part of the mixed gases passing through the water tank is entering the sensing module with a consistent sampling flow rate. The data processing module employs an algorithm to adjust the output values from the hydrogen sensor, accurately reflecting the true hydrogen concentration in the accident environment. The accuracy of hydrogen concentration monitoring in a severe accident environment is assessed based on the test results obtained from the experimental matrix.

The generation of combustible gases (H₂ and CO) during a severe accident (SA) and their potential accumulation in the containment atmosphere could threaten the containment integrity and/or safety components in case of uncontrolled combustion. The AMHYCO project (2020-2025), funded by the European Commission, aims to enhance the understanding of H₂/CO combustion risk within the containment of a nuclear power plant, particularly in the late phase of a severe accident, to revise the management of combustible gas risk.

This work, performed in the frame of AMHYCO, explores the impact of passive autocatalytic recombiners (PARs) performance on SA progression, and particularly their interaction with other safety systems (i.e., sprays and fan-coolers). Two Western PWR scenarios (a double-ended guillotine LOCA and an SBO) were simulated with the MELCOR 2.2 code. In the LOCA scenario, steam concentration is strongly reduced shortly after the initiating event by the automatic spray actuation. The suppression of steam promotes the formation of flammable gas mixtures in the ex-vessel phase. Parametric cases showed that cooling systems' unavailability or deactivation could reduce combustion risk. Contrarily, the SBO accident initially evolves at high pressure with a high steam content in the containment. In this sequence, a late spray operation significantly affects the gas mixture's flammability. In both sequences, the oxygen depletion by the PARs operation leads to containment inertization in the late phase of the accident. As a future step, CIEMAT will launch a calculation campaign to assess how uncertainties may impact the insights gained through best-estimate analyses.

The mitigation of the hydrogen risk with passive auto-catalytic recombiners (PARs) is state-of-the-art in nuclear power plants with water-cooled reactors. In the ex-vessel phase of a severe accident, the operation of PARs faces several challenges. While hydrogen is continuously released from the interaction between molten corium and concrete, carbon monoxide is also produced, along with other gases. Inside the PAR, hydrogen and carbon monoxide compete for the available oxygen, which is continuously consumed. As a consequence, the performance of the PAR in terms of recombination rates and overall efficiency decreases. In order to enable a realistic assessment of the availability and performance of the measures to control combustible gases, numerical models developed for PAR operation during the in-vessel phase need to be enhanced towards these boundary conditions.
The REKO-DIREKT code was developed to describe the operational behavior of PARs with a mechanistic approach. This model describes heat and mass transfer processes inside the catalyst section in detail in order to simulate PAR performance and conversion efficiency. Based on an experimental program which was part of the European AMHYCO project, REKO-DIREKT was further developed to cope with the ex-vessel conditions of a severe accident.
The paper describes aspects of the code development work and gives examples for the code validation based on the existing database.

The Japan Nuclear Regulation Authority recognizes that the hydrogen explosion at Fukushima Daiichi Nuclear Power Plant Unit No. 3 originated not on the operating floor but on the lower level of the reactor building. This study aims to obtain knowledge on hydrogen behavior by examining analytical conditions under severe accident scenarios to develop an evaluation method for the retention and diffusion behavior of hydrogen leaked into the lower level of the reactor building, which represents a typical BWR plant in Japan. Based on plant walk-down data and the results of safety analysis evaluation of the actual plant, the dimensional shape and heat transfer characteristics of the floor area where hydrogen may leak, and the fluid characteristics of the leaking gas, were organized. An analytical model was developed using representative parameters as basic conditions. Sensitivity analysis of various parameters showed that the height from the leak point to the ceiling and the horizontal distance to the ceiling cavity were highly sensitive to the hydrogen concentration in the ceiling cavity. The hydrogen concentration increased as the vertical distance from the leak location to the ceiling decreased, and as the horizontal distance to the ceiling cavity decreased. By contrast, other parameters, such as the temperature of the leaking gas, had little effect on the hydrogen concentration in the ceiling cavity. The results of the sensitivity analysis indicate that these are the three main factors that increase the hydrogen concentration in the ceiling cavity.

During a severe accident, hydrogen distribution in a containment building and characteristics of hydrogen depletion by PARs differ depending on the thermal-hydraulic behaviors occurring in the containment. Various pressure control systems are installed in the containment building to prevent overpressure in severe accident conditions. Representative systems include a spray, a fan-cooler (RCFC: reactor containment fan-cooler), a filtered containment venting system (FCVS), and a passive containment cooling system (PCCS). The containment pressure control system ensures the integrity of the containment building by maintaining the containment pressure lower than the design pressure in a severe accident condition. However, during the operation of this pressure control system, the effectiveness of the hydrogen control system and the hydrogen safety in the containment building must be ensured. This study intends to experimentally evaluate the hydrogen removal characteristics of a PAR when pressure control systems such as RCFC, and PCCS are operating. The following were obtained from the experiment. In the PAR-PCCS experiments, the hydrogen removal rates of the PAR show a similar value to the correlation even during the PCCS operation, so it seems that the PCCS has little effect on the PAR operations. It is judged that the operation of the RCFC does not hurt the removal of hydrogen from the PAR through the evaluation experiment of the PAR performance according to the operation of the fan cooler.

The paper outlines ongoing research in a national funded joint project, applying machine learning methods to predict late-phase phenomena observed during severe accidents. The aim is to produce resource-efficient simulations that improve the understanding and predictive capabilities for these late-phase phenomena. Emphasis is on the long-term coolability of debris beds in the vessel, its remelting and possible relocation in the cavity as ex-vessel debris bed. For this purpose, a machine learning model is intended to be integrated to a PSS inhouse version of AC² program package, developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, to make the complex calculation of heat transfer and dryout heat flux more efficient.
The aim is to develop an artificial neural network as a machine learning model. Several different network types are initially developed for this purpose, for example a physically informed neural network (PINN). As work progresses, these models should become more complex and be able to provide a fast and reliable prediction of the debris bed coolability for as many different boundary conditions as possible. According to the current status, the tested network types are already able to produce good preliminary results. They are able to predict the dryout heating power, selected as the output variable, to a good approximation. The training data are generated using the COCOMO-3D detail code from the Institute for Nuclear Energy and Energy Systems (IKE) at the University of Stuttgart.

The recent developments in data-driven methods have paved the way to new methodologies to provide accurate state reconstruction of engineering systems; nuclear reactors represent particularly challenging applications for this task due to the complexity of the strongly coupled physics involved and the extremely harsh and hostile environments, especially for new technologies such as Generation-IV reactors. Data-driven techniques can combine different sources of information, including computational proxy models and local noisy measurements on the system, to robustly estimate the state. This work leverages the novel Shallow Recurrent Decoder architecture to infer the entire state vector (including neutron fluxes, precursors concentrations, temperature, pressure and velocity) of a reactor from three out-of-core time-series neutron flux measurements alone. In particular, this work extends the standard architecture to treat parametric time-series data, ensuring the possibility of investigating different accidental scenarios and showing the capabilities of this approach to provide an accurate state estimation in various operating conditions. This paper considers as a test case the Molten Salt Fast Reactor, a Generation-IV reactor concept, characterised by strong coupling between the neutronics and the thermal hydraulics due to the liquid nature of the fuel. The promising results of this work are further strengthened by the possibility of quantifying the uncertainty associated with the state estimation, due to the considerably low training cost. The accurate reconstruction of every characteristic field in real-time makes this approach suitable for monitoring and control purposes in the framework of a reactor digital twin.

In the event of a severe accident in a nuclear power plant, accident prediction using artificial intelligence (AI) has gained attention as a promising Accident Management Support Tool (AMST). A notable approach is the development of surrogate models for accelerated accident prediction through deep learning-based supervised learning. Such models alleviate the computational complexity of severe accident analysis codes by training on data generated from the codes, significantly reducing the computational costs.

However, surrogate models often present structural challenges, leading to low-resolution predictions and increased uncertainty, hindering effective decision-making for operators. This issue contradicts the essential requirements for AMST reliability. Structural issues arise from low temporal resolution and information loss during data preprocessing for training, limiting the model's accuracy due to cumulative computational errors in time series forecasting. Consequently, using surrogate models to predict thermal-hydraulic variables with refined time resolution during accident progression can yield unreliable results.

To address these challenges, this study aims to develop an auxiliary surrogate model to support accident prediction by identifying time varying patterns in the accident prediction data. This model is designed to predict the onset time and amplitude of physical variations, enhancing the accuracy and reliability of surrogate-based predictions during severe accidents.

This work is performed in the frame of the EU-funded project ASSAS (Artificial intelligence for Simulation of Severe AccidentS), which aims at developing a basic-principles severe accident simulator for a generic PWR-1300MW, by replacing models from ASTEC (severe accident code developed by IRSN) with machine-learning surrogate models. PSI’s tasks address essentially the CESAR module (thermal-hydraulic solver) in the primary and secondary circuits. This paper proposes a surrogate model able to reproduce the thermal-hydraulic behaviour of the reactor cooling system (RCS) during the early phase of a Station Blackout (SBO), i.e., until hydrogen generation, with a significant speed-up factor.

A suitable training dataset must be generated. A base case scenario is considered, involving a SBO without any safety measures until the onset of core oxidation. From this base case, various calculations are performed by depressurizing remotely the primary circuit at different times, followed by the recovery of emergency water injection, also at different times. From these ASTEC calculations only the variables needed for the surrogate model development are extracted. These include state variables for each control volume within the RCS domain, boundary conditions, and additional variables that provide information about the overall evolution of the accident and are useful for Machine Learning. The surrogate model is expected to compute each time-step, like ASTEC does, while also accounting for user decisions interactively during the accident simulation. The Machine Learning methods considered in this work are based on artificial neural networks, and more specifically recurrent neural networks, which are commonly used for time-series.

Small Modular Reactors (SMRs) are gaining increasing attention due to their enhanced safety features, flexibility, and scalability. Ensuring timely and accurate assessments of break sizes during break accidents is crucial for maintaining the safe and reliable operation of SMRs. However, current methods for evaluating break sizes mainly rely on the personal judgment of operators, which often fail to meet the speed and accuracy requirements in high-risk, time-sensitive situations. This limitation can hinder effective decision-making and risk management. Recent advancements in artificial intelligence (AI) have accelerated the development of data-driven methods for break size prediction, demonstrating significant potential for improving operational reliability. Machine learning models, particularly those with interpretability features, can provide real-time, data-driven predictions of break sizes, offering a faster and more accurate alternative to traditional methods. Furthermore, the interpretability of these models can foster greater trust in AI systems, particularly in safety-critical environments such as nuclear reactors. This study investigates Direct Vessel Injection (DVI) break accidents, utilizing explainable artificial intelligence (XAI) and hyperparameter optimization techniques to develop predictive models for break size. The results demonstrate that these models enable rapid and accurate prediction of DVI break sizes based on actual operational parameters. The findings of this research provide valuable insights for developing break size prediction models base on SMRs, contributing to improved safety and operational efficiency.

This research focuses on the safety evaluation of a design extension condition involving multiple steam generator tube rupture (MSGTR) scenarios. A series of operator actions are proposed to mitigate the accident, including depressurization, auxiliary spray operation, and steam generator blowdown. The efficacy of these actions is evaluated under various uncertainties using the best estimate plus uncertainty (BEPU) approach through RELAP5/DAKOTA coupling. The generated ensemble of system responses is used to develop an AI-based prediction model. Tools of explainable artificial intelligence, specifically a combination of attention mechanisms, gradient-based attribution, and parameter interaction analysis, are implemented to examine the model's decision-making process. This framework reveals phase-specific patterns and dynamic shifts in parameter relevance as the accident progresses through different stages–from initial break flow and pressure response, through various operator interventions, to final stabilization. The analysis quantifies the coupling between primary and secondary systems, particularly during critical phases of depressurization and cooldown, while demonstrating the model's adherence to established thermal-hydraulic principles. The result highlights the AI model's general alignment with established thermal-hydraulic principles, suggesting its potential for integration into nuclear safety management, provided its transparency and interpretability continue to be rigorously validated.

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.