Conference Agenda

Rod bundle spacer grids are known to have important, possibly dominant effects on thermal-hydraulic behavior during core uncovery. Spacer grids enhance mixing and convective heat transfer downstream of the grid Droplet breakup at a spacer grid increases interfacial area and the de-superheats steam benefiting heat removal from uncovered rods. Because they are unpowered the spacer grids can also rewet much easier than surrounding fuel rods. The liquid film on a wet grid acts as a source for droplet entrainment immediately downstream of the grid which further increases interfacial area and steam de-superheat.

The Rod Bundle Heat Transfer Facility (RBHT) data provide a unique opportunity to determine conditions for spacer grid rewet during reflood. Each spacer grid in RBHT has one or more wall-mounter thermocouples that indicate if and when the grid rewets. Tests ranged in RBHT from low flooding rates (0.5 cm/sec) to 15 cm/sec. Previous studies have shown that spacer grids can rewet well ahead of the quench front on the heater rods. However, no systematic study has been done to characterize when and which spacer grids rewet. The effects of wetted grids on the droplet dynamics and two-phase heat transfer downstream of the grids have not been seriously investigated.

In this paper results from several RBHT tests are presented and discussed with an emphasis on spacer grid rewet and the challenges in simulating the rewet process and effects on transient reflood behavior.

The global energy landscape is undergoing a nuclear renaissance, driven by rising energy demands, economic growth, and the computational needs of artificial intelligence (AI) and machine learning (ML). With around 65 reactors under construction across 15 countries—and more planned, including in nations new to nuclear energy—the industry is advancing swiftly. While most new reactors rely on traditional light water reactor (LWR) designs, a growing number embrace advanced concepts like heat pipe, gas-cooled, liquid metal-cooled, and molten salt reactors. These “new” advanced designs, grounded in mid-20th-century thermohydraulic (TH) research, face modern demands for safety, economic viability, and regulatory compliance. A key challenge is the reliance on outdated TH experimental data, which, limited by past techniques’ resolution, lacks the spatial and temporal detail to capture next-generation reactors’ complex flow physics, hindering validation of simulation tools like CFD. In this paper, we present examples of high-resolution thermohydraulic experiments designed to support the development of advanced nuclear reactors. These experiments address key challenges in model validation by transforming raw measurements into trusted datasets. Using cutting-edge diagnostics and instrumentation, they deliver precise, high-fidelity data that capture critical flow phenomena under relevant conditions. The resulting datasets are used to validate CFD and multi-physics simulation codes, reduce modeling uncertainties, and support improved physical understanding. By strengthening the predictive capability of simulation tools, these experiments contribute to refined reactor designs, optimized performance, and more efficient regulatory licensing—ultimately enabling the safe and effective deployment of next-generation nuclear technologies.

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

Dispersed flow film boiling (DFFB) is a key flow regime that affects fuel rod integrity during emergency core cooling system (ECCS) injection phase in large-break loss of coolant accident in light water reactors. However, the existing experimental data in the DFFB regime have limitations in the important parameters that were measured, such as the void fraction and vapor superheat. These limitations in the DFFB data lead to remarkable uncertainties in the models/correlations developed for the wall convective heat transfer. To improve our understanding of and obtain experimental data for the wall heat transfer characteristics in the DFFB regime, a series of quasi-steady-state DFFB experiments was performed in the Post-CHF Heat Transfer (PCHT) test facility, over flow conditions of mass flux from 60 to 150 kg/m²-s, and pressure from 1.38 to 4.14 bar. The obtained wall convective heat transfer coefficient shows a transition in its trend with the thermodynamic equilibrium quality due to the transition of the dominant heat transfer mechanism from interfacial heat transfer to the vapor convection. The transition mechanism was confirmed by analyzing the computed vapor Reynolds number and the measured void fraction by an X‑ray radiography system. Based on the collected data, a new wall convective heat transfer correlation was developed by applying the Reynolds analogy to consider the variation of the interfacial area concentration with the droplet size. The newly proposed correlation successfully demonstrated its improved predictive capability compared to the existing models/correlations for a wide range of DFFB wall heat transfer data.

Best Estimate Plus Uncertainty (BEPU) methodologies are gradually establishing themselves as the favored way to perform Deterministic Safety Assessments (DSA) of Nuclear Power Plants (NPPs), because they account for uncertainties in plant states and physical behaviors while employing accurate-to-reality (best estimate) simulation codes. This work is performed in the context of the ATRIUM project, which seeks to establish best practices and standardize the BEPU process to make the results consistent, minimizing user effect. The approach taken in this project is to focus on specific phenomenology that is crucial to the progression of the desired transient (a Loss of Coolant Accident), and work with Separate Effect Tests (SET) experiments to find adequate uncertain parameters and their PDFs to propagate.

The target phenomenology for this study is that of post-Critical Heat Flux (CHF) heat transfer, concretely film boiling. Simulations of the Becker film boiling experiments were performed using RELAP5 to validate its suitability and identify influential parameters. Initial tests showed that RELAP5 overestimated burnout quality, resulting in inaccurate axial temperature profiles. Also, a sensitivity analysis of several RELAP parameters and initial conditions showed they didn’t influence the results.

To address this, RELAP5’s source code was modified to introduce and externalize new parameters for better control over film boiling modeling. Consequently, a sensitivity analysis was carried out on this new group of parameters, helping measure their effect and aiding in fine-tuning them to improve RELAP5’s predictions. This enhanced flexibility demonstrates a promising approach for more accurate RELAP5 analyses of complex post-CHF phenomena.

Prediction of the quench temperature is one of the most important parameters for the accurate estimation of the accident progress in light water reactors (LWR). In this study, a single rod flow quench experiments were conducted in the subcooling range of 0 to 40 K and 600-1100 °C initial cladding temperature conditions. Measurements of quench temperature were made at four levels of elevation from the bottom of the furnace. Based on these comprehensive tests, unique quench temperature trends were identified comprising two distinct regimes: (i) increase of quench temperature with increasing subcooling and initial cladding temperature with less effect of elevation in the high subcooling and low initial cladding temperature regime and (ii) constant quench temperature independent of subcooling and initial cladding temperature with a large effect of elevation in the low subcooling and high initial cladding temperature regime. The criteria delineating the two regimes was determined as a function of both subcooling and initial cladding temperature. A thorough analysis of cooling rate during the film boiling regime and quench front velocity was performed to develop a quench temperature correlation to better understand this new finding on quench temperature behavior. The study covers a critical gap in literature where high initial cladding temperatures (over 800 °C) under wide range of subcooling for such experiments have not been typically explored.

Quenching of hot vertical wall with a falling liquid film is an important thermal-hydraulic process to ensure the safety of nuclear reactors even during emergency situations. In this work, to develop a reliable model to predict the propagation velocity of the quench front, the temperature distribution of heat transfer surface during quenching was measured using a high-speed infrared camera. A silicon wafer that is transparent to the infrared ray was used as the hot wall, and the initial wall temperature, the wall thickness, the cooling liquid temperature, and the liquid flow rate were changed parametrically. The main heat transfer mechanism from the wall to the liquid film near the quench front was found to be the nucleate boiling. The heat transfer coefficient profile derived from the measured temperature distribution was therefore correlated using widely accepted heat transfer correlations such as Zuber's correlation for pool boiling CHF (Critical Heat Flux) and Rohsenow’s correlation for the pool boiling HTC (Heat Transfer Coefficient). The calculated quench velocity was in good agreement with the experimental data not only for the silicon wafer but also for the copper and zirconium walls of different thermal properties.

Despite their importance and ubiquity, high void fraction two-phase flows are notoriously difficult to probe optically. With a significant number of dense scattering bubbles, images become rapidly corrupted by intensely scattered photons, resulting in occlusion, loss of definition, and errors in size and motion estimation. Information lost through these detrimental effects can be recovered if scattered photons are excluded entirely. However, this corrupting light appears femtoseconds to picoseconds after initial illumination, depending on bubble sizes. Achieving such short measurement times is not possible by conventional electronic means. Through non-linear optical phenomena, including the Optical Kerr Effect, images can be acquired with picosecond and sub-picosecond exposure times and gated in 33 femtosecond intervals.
Using semiconductor and liquid crystal-mediated Kerr gating, static and dynamic bubbly flow scenes have been temporally decomposed, capturing only useful imaging light. In doing so, the definition of bubbles is preserved, reducing uncertainty in size determination, even in optically thick scenes. Furthermore, as the spatial scattering profile is time-dependent, experimentally capturing the complete scattering history allows bubble size measurement. Applying this technique to a newly built, high void fraction bubble column, proof of concept is shown for simultaneous tracking and measurement of size and deformation. With continued development, ultrafast imaging may significantly improve multiphase imaging and provide insight into ultra-high-speed transient events.

In order to further decarbonize energy uses, numerous compact and small-scale nuclear systems are being developed worldwide. In this context, there is a growing interest in conducting research on two-phase flows in mini-channels, as understanding the behavior of these flows is essential for optimizing the heat transfer processes and enhancing the efficiency of these advanced nuclear systems. Among the thermal-hydraulic issues identified in this field is the frictional pressure drop under two-phase conditions. The present study addresses those issues by means of an experimental investigation of the two-phase frictional pressure drop which was carried out at a millimetric-scale. Those laboratory experiments were conducted using a set of two mini-channels of different length, with an inner diameter of 1.38 mm and arranged horizontally, under controlled conditions to measure the pressure drop as a function of imposed phasic flow rates. Demineralized water was used as the working liquid. In order to simulate an adiabatic two-phase flow, air was injected within the liquid at the inlet of the mini-channels. Phasic mass flow rates were imposed up to 9 g/s and 0.14 g/s, yielding maximum Reynolds numbers of 21,000 and 8,000, respectively for the liquid and gas phases. A dimensionless two-phase pressure drop was calculated from the acquired data and compared with the most recommended model for pressure drop in mini-channels to date. This model has proven incapable of reproducing the experimental data. This highlights the need to improve the predictability of pressure drop models in mini-channels.

This study presents a miniaturized impedance sensor probe designed for two-phase flow measurements, with significant implications for nuclear reactor systems. Positioned between Electrical Resistance Tomography (ERT) and optical/resistive needle probes, the sensor is based on closely spaced parallel needle electrodes to accurately measure local flow parameters, including void fraction, interfacial velocity, and bubble size. Its design allows for high spatial resolution without requiring interaction with the bubble surface, a key advantage over conventional local intrusive probes.

Numerical simulations to assess the electric field distribution across various electrode configurations were employed for two critical purposes. First, assess the impact of bubbles passing outside the primary measurement zone (inter-electrode area), which is essential for understanding how off-axis flow phenomena affect measurement accuracy. Second, the simulations provided insights into the signal processing requirements by generating simulated sensor outputs. These outputs were a key aspect for refining the algorithms used to extract key flow parameters from the sensor data.

Experimental validation, including high-speed imaging and comparisons with resistive probes, confirmed the sensor's capability to detect smaller bubbles and continuously track two-phase flow changes in real-time. The combination of its robust design, high spatial resolution and temporal resolution makes this sensor a promising alternative to existing technologies, suited for applications in nuclear reactor coolant monitoring, where precise control over multiphase flows is essential for ensuring system safety and performance.

Aerosols play a critical role in the transport of radioactive products within nuclear reactors. During severe accident scenarios, high-temperature and high-pressure coolant sprays can lead to complex temperature distributions within the containment, influencing the thermophoretic transport, evaporation, condensation, and coalescence of aerosols. Aerosol measurement technologies have evolved significantly over the years, leading to the development of diverse methodologies, including single-point/full-field, sampling/in-situ, and intrusive/non-intrusive approaches. For instance, laser-based particle visualization has been widely employed to study aerosol particle dynamics. However, temperature field measurements during aerosol transport remain rarely reported in the literature. This study introduces a novel non-intrusive method capable of simultaneously capturing the temperature field during aerosol transport and visualizing the motion of aerosol particles. The Background Oriented Schlieren (BOS) method is employed to obtain the temperature fields of aerosol particle flow through a controlled temperature gradient in a visualized channel. Experimental results demonstrate that this method can accurately obtain the velocity and temperature fields within the measurement domain, with uncertainties less than 5% for the temperature field. Additionally, this study quantitatively analyzed the influence of aerosol introduction on BOS measurements through comparative experiments. The results indicate that light scattering caused by the aerosol particles has no significant effect on the BOS measurement outcomes.

CATHARE is the French thermal-hydraulic code used for nuclear reactors safety analysis. Its reflooding module has been extensively validated for Large Break Loss-Of-Coolant Accident (LB-LOCA) scenarios. However, as Intermediate Break LOCA (IB-LOCA) studies are becoming more and more frequently, the validation of the CATHARE reflooding module needs to be extended to higher pressures than those encountered during LB-LOCA core reflooding.

The PERICLES experimental program at CEA primarily aims to improve the understanding of core thermal-hydraulics during the reflooding phase of a Pressurized Water Reactor (PWR). The test section consists of an insulated stainless-steel shroud, containing 368 full-length electrically heated fuel rod simulators (FRSs) and 25 stainless steel guide tubes arranged in a 17 × 17 geometry. The PERICLES facility has been considerably used for validating the CATHARE code. The experimental program includes 15 high-pressure (10 to 60 bar) reflooding tests which have not previously been used for CATHARE validation.

This paper first introduces the CATHARE code, the reflooding module, and its adaptation to pressures above 6 bar. Then, it presents a comparison between the PERICLES experimental data and the CATHARE computation results. The CATHARE code demonstrates a good prediction of the reflooding tests. Activating the reflooding module of the CATHARE code allows better predictions compared to results without the module's activation.

This paper will conclude with a discussion on the limitations of the presented validation and the need of further experimental data on high-pressure reflooding.

Direct Numerical Simulation (DNS) is regarded as an accurate and reliable approach to generate high-resolution data. However, due to the high computational costs involved, DNS investigations are typically limited to simple geometries and scaled-down conditions. In addition to providing an insight into the flow physics, fully-resolved DNS data is often used as a valuable reference for development, validation and improvement of Computational Fluid Dynamics (CFD) models. It is shown in this study, Under-resolved DNS (UDNS) based on spectral element methods (SEM) offers a more cost-effective alternative for generating high-quality reference data, while still providing sufficiently accurate flow statistics with significantly fewer degrees-of-freedom. Moreover, calculating the statistical quantities of the flow using UDNS does not require the additional effort involved in including sub-grid modeling contributions.

The resolution and accuracy of DNS and UDNS simulations are evaluated for an internally-heated natural convection flow at Rayleigh number of 10¹¹. Three UDNS simulations are performed at progressively coarser grid resolutions, thereby reducing the computational costs. It is shown that high-accuracy DNS can only be achieved at a grid resolution criteria of Δ/η_k ~ π. The UDNS approach is shown to achieve sufficiently accurate results for the mean and RMS velocity and temperature fields, as well as the turbulent kinetic energy budgets. Such UDNS statistical data can serve as a cost-effective alternative for reference for the validation of engineering CFD models. Furthermore, high-quality reference data can be obtained at more realistic flow conditions at higher Reynolds or Rayleigh numbers, where fully-resolved DNS may be infeasible.

Direct Numerical Simulations (DNS) are considered an accurate and reliable approach to generate high-resolution data. However, due to higher computational costs involved, DNS is generally performed in scaled and/or simpler geometry, which is representative of the real industrial-scale scenario. In order to represent the flow and heat transport between an array of fuel rods in a reactor core, simulations are conventionally performed in a smaller computational domain with periodic boundaries – often limited to a single interstitial subchannel space or domain around a single pin.

The present study examines DNS results for several different configurations – a periodic domain around a single pin, a single subchannel space, and arrays of 2×1 and 2×2 subchannels. It is shown quantitatively that the proximity of the periodic boundaries in the smaller domains significantly alters the flow physics. The periodic boundaries of the smaller domains, being highly-correlated, cannot faithfully represent the large-scale flow interaction across subchannels. Higher-order flow statistics, which may be used for development and validation of lower order models, are shown to be significantly affected by the periodic boundaries of the smaller domain. The distribution of wall shear stress and Nusselt number around the pin are also seen to be affected. A comparison of secondary flows in the different geometry representations is also presented. Flow pulsations in the narrow gap, typically associated with low P/D ratios, are also observed for the present P/D ratio of 1.326. The frequency of these pulsations is also shown to be affected by the size of the domain.

The supercritical water-cooled reactor is one of the proposed designs under further
study and development by the Generation IV International Forum (GIF). Like any
emerging technology, it presents technical challenges that must be resolved and better
understood. This paper addresses one of these challenges, from a thermal hydraulics
perspective, by focusing on the effect of idealised surface roughness on flow and heat
transfer in mixed convection of supercritical water.
To investigate this phenomenon, we perform six DNS simulations for fully developed,
asymmetrically heated channel flows. Three cases feature channel walls (top and
bottom) tiled with pyramids to represent an idealized surface roughness of Rt (peak-to-
valley height) 0.05δ (where δ is the half-channel height), while the remaining three have
smooth walls for comparison. We examine three flow conditions: (i) vertical flow with
mixed convection, (ii) stably-stratified mixed convection horizontal flow, and (iii) forced
convection. Constant temperature boundary conditions are applied at the channel walls,
with the hot side having a temperature greater than the pseudocritical temperature and
the cold wall temperature smaller than it. The temperature difference between the walls
is kept at 15 K. The aim of this work is to study the impact of roughness on heat transfer
deterioration and enhancement. Comparisons are made between mixed and forced
convection flows in both vertical and horizontal orientations, and the DNS data provided
will be useful for benchmarking.
The simulations employ the open-source DNS solver CHAPSim 2.0, with the prescribed
surface roughness modelled using an Immersed Boundary Method (IBM). This method
allows for the modelling of complex geometries by introducing a body force to the solid
nodes.

CFD codes used in the nuclear industry require an extensive validation, over a large range of thermal hydraulic conditions. In the context of PWR scenarios with shutdown of the primary pumps, buoyancy-affected flows in the core rod bundle can be encountered at low flow rate. However, scarce experimental data are available for this type of flow.

To address this issue, a Wall-Resolved LES (quasi-DNS using high-order schemes) of an upward flow within a heated rod bundle subjected to lateral power skews has been performed by Vicente Cruz et al.[1]. This numerical experiment of a complex turbulent mixed convection flow using a Boussinesq approximation, exhibiting cross flows, a mixing layer and a local re-laminarization, provides a highly accurate database for code validation.

In the present work, following the recommendations by Benhamadouche[2], the same configuration is investigated using code_saturne[3]. The predictions show good agreement with the reference database which demonstrates the capability of this in-house industrial code to carry out high-fidelity LES computations for this type of configuration.

[1] R. Vicente Cruz et al., “Numerical investigation of the mainly axial flow in mixed convection regime within tube bundles”, Proceedings of the 18th UK Heat Transfer Conference (2024)

[2] S. Benhamadouche, “On the use of (U)RANS and LES approaches for turbulent incompressible single phase flows in nuclear engineering applications”, Nuclear Engineering and Design, 312, pp. 2–11 (2017)

[3] code_saturne is an open-source incompressible CFD code developed by EDF (https://www.code-saturne.org/cms/web/).

Twisted elliptical tubes are a proposed heat transfer enhancement (HTE) for use in Molten Salt Reactors (MSRs) due to their enhanced thermal performance compared to plain tubes. The twisting ellipsoid geometry causes the fluid to swirl as it passes over the tubes, enhancing mixing and inducing turbulence at lower Reynolds numbers. This effect increases the overall convective heat transfer; however, this is at the cost of increased frictional pressure losses. In this study, the Nek5000 computational fluid dynamics (CFD) code is used to perform Large Eddy Simulations (LES) of external flow around twisted elliptical tube bundles. Simulations were performed for three different tube cross-sectional aspect ratios (maximum versus minimum diameter), Reynolds numbers of 1,000 and 7,000, and Prandtl numbers ranging from 1-25. The aim of this study is to further characterize the heat transfer phenomena seen around twisted elliptical tubes when varying the tube aspect ratio, which is a dependency that has not been historically accounted for. Previous work by the authors has investigated the dependency of cross-sectional aspect ratio (AR=1.1 – 2.1) and below unity Prandtl numbers (Pr=0.001 – 1) for this geometry, and this study aims to extend this work by investigating above unity Prandtl numbers and varying Reynolds numbers.

Thermal radiation plays a crucial role in heat transfer for next-generation nuclear reactors due to the high operating temperatures and reliance on natural convection. In our previous work, we successfully implemented the P1 approximation for thermal radiation in Nek5000/NekRS, a CFD code based on the Spectral Element Method. The implementation was verified against both numerical data and analytical solutions. However, that work focused solely on the fluid domain.

In this paper, we extend the P1 model to the solid domain by integrating it with the Conjugate Heat Transfer model in Nek5000/NekRS. As before, we validate our implementation with reference numerical solution on simple geometries. Then, we apply this approach to the pebble bed case with 1,568 pebbles under salt flow cooling, which was introduced in our previous work, now including the solid domain of pebbles. To ensure accuracy, we conducted parallel simulations using OpenFOAM for comparison.

The presence of a large-scale interface between the two phases presents a modeling challenge related to turbulence in a sense that at a large-scale interface the lighter phase sees the heavier phase like a solid wall. In literature, a widely used strategy is to add a damping source term to the turbulence dissipation equation. These source-term based approaches require that a parameter be tuned based on grid and/or problem. In the present work, a novel approach is proposed where the use of solid wall-function based turbulence treatment is done within the framework of LSI model implemented in Simcenter STAR-CCM+. This is achieved by using a large-scale interface toolkit, which provides the location of the large-scale interface as well as the cell-center to large-scale interface distance; enabling creation of a stencil around the large-interface cell to apply wall-function base damping treatment.

The effectiveness of the new wall-function type interface turbulence damping method is demonstrated in this work though a turbulent air-water co-current stratified flow, studied experimentally by Fabre et al., (1987). The wall-function type treatment shows minimal dependence on grid refinement as well as on the cell aspect ratio when comparison is performed for the full developed velocity profile against the experiment.This is not the case for source term based approach. The observations made using the velocity profile are corroborated for the pressure drop values as well. In addition, the novel approach is shown to not need any problem-based or grid-based tuning.

Two-phase flows are of great interest to the nuclear power community, both in normal operation for coolant systems, and in accident scenarios. Computational multiphase fluid dynamics (CMFD) simulations are a promising tool for detailed analysis of these systems. However, since most CMFD models are based on experiments and analysis of vertical two-phase flows, significant limitations are evident when horizontal bubbly two-phase flows are considered, which have entirely different hydrodynamics arising from the large density difference between phases. In this work, a new experimental database of relative velocity for horizontal bubbly flows is established utilizing the existing 25.4 mm test facility at Purdue University. Local void fraction, gas velocity, and bubble diameter are measured with four-sensor conductivity probes, while local liquid velocity is measured with a Pitot-static probe. This data is used to evaluate the existing CMFD models in ANSYS CFX, demonstrating the problems with the existing models, especially with predicting relative velocity. A recently proposed relative velocity model is implemented in CFX, which improves both the relative velocity and void fraction prediction. Qualitatively, the void fraction location and shape are greatly improved, peaking further away from the wall with an elliptical shape that better agrees with the experimental data. The CMFD predicted area-averaged void fraction matches the experimental data about 10% better with the new model, while the relative velocity is improved by 30%.

Turbulence induced vibration of fuel rods can lead to mechanical wear, which can be responsible for safety issues and significant maintenance costs in Nuclear Power Plants (NPPs). In the context of the “Gathering expertise On Vibration ImpaKt In Nuclear power Generation” (GO-VIKING) European project, fluid structure interaction (FSI) methods are developed and assessed on simpler geometries. Hereby, two-way coupling simulation, of a cantilever rod under a turbulent axial flow have been performed at Reynolds number 21 200. Computational fluid dynamics (CFD) with finite volume method and wall-resolved large eddy simulation (WR-LES) is used for the flow, along with an in-house Euler-Bernoulli beam equation solver for the rod. Free vibrations tests have been performed for the calibration of the mechanical damping of the beam and for validation purposes. The beam motion is accounted for using an Arbitrary Lagrangian Eulerian (ALE) method. The meshes are made of 144 million hexahedra for the flow and 3800 grid points for the beam, respectively. The beam motion and the flow statistics are both investigated; comparisons are drawn with experimental results from the literature. The theoretical and experimental natural frequencies are recovered by the simulation. The R.M.S of the tip displacement of the beam is lower in the simulation than in the experiment. However, this may be due to both the experimental uncertainties at this Reynolds number and a misalignment of the rod in the experiment. Apart from the dissymmetry due to the misalignment, there is quite a good agreement regarding the flow statistics.

Natural convective heat transfer is a prominent research topic in the nuclear energy field and is crucial for the design of passive safety systems. To study the impact of compressibility-induced non-Oberbeck-Boussinesq (NOB-II) effects on natural convection, we conduct lattice Boltzmann simulations of square cavity natural convention in a perfect gas, incorporating the multiple-relaxation-time force model and pseudopotential force. The findings indicate that for a given Rayleigh number (Ra), the Nusselt number (Nu) increases as NOB-II effects strengthen and the Reynolds number (Re) decreases as these effects intensify. This implies that NOB-II effects lead to heat transfer enhancement and convection suppression. The underlying mechanism is as follows (taking the hot fluid as a representative case): under NOB-II conditions, the compression work term absorbs heat from the hot fluid near the central region of the hot wall, resulting in a steeper temperature gradient and a thinner temperature boundary layer near the hot wall. Consequently, the local Nusselt number increases and overall heat transfer is enhanced. Simultaneously, the reduction in the thickness of the temperature boundary layer causes a decrease in the buoyancy difference, ultimately leading to convection suppression. Furthermore, new scaling laws of Nu-Ra and Re-Ra considering NOB-II effects are proposed, with an average error of less than 5%. This study deepens the understanding of natural convection and offers theoretical support for the thermal-hydraulic and safety analyses of advanced reactors.

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

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

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

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

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

Severe accident analyses in nuclear power plants are highly complex and models applied in simulation codes are often derived based on limited amount of available experimental data. Evaluating the accuracy and uncertainty of such models provides valuable information for safety analyses, as well as further development needs and helps the priorisation of experimental resources. Consequently, there has been a growing interest in the recent years in uncertainty quantification and BEPU analyses in the context of severe accident analyses.

This paper investigates a postulated cold leg break combined with a station blackout scenario in a generic German PWR using the AC2 code package, developed by GRS. Both the reactor cooling circuit and containment are modelled in detail and the scenario is analysed starting from normal operational conditions up to the evaluation of the source term into the environment.

Based on the best estimate simulation an uncertainty and sensitivity analysis is performed. The selected uncertain input parameters are propagated through the model and the 95/95 uncertainty ranges are determined based on Wilks` findings. Furthermore, the Spearman Rank Correlation Coefficient is derived as sensitivity index. This allows to characterize both the variation in model responses and identify important uncertain parameters.

In 2019 was launched the Horizon-2020 MUSA project with the aim of reviewing the uncertainty sources as well as defining Uncertainty Quantification methodologies for assessing Severe Accidents (SA) scenarios. Energy Software Ltd. (ENSO) contributed to the Working Package 5 “reactor applications” with the simulation of a long-term Station Black Out (SBO) occurring in a Gen-II PWR. The Uncertainty Analysis (UA) was carried out with RELAP/SCDAPSIM/MOD3.4 (RS3.4) and its IUA2.0 module, using the Wald multi-variable form of the Wilks’ equation. A group of 20 input parameters and 19 Figures of Merit (FOM) were selected to assess the uncertainty propagated to the Source Term (ST) released to the containment in an in-vessel simulation. To improve the ST capabilities of the code, a Fission Product Transport model was implemented into RS3.4. The relevant outcome of ENSO contribution was the estimation of tolerance regions for the Fission Products, Gases and Debris materials released at the vessel failure. Such outputs can be then used as initial conditions and/or probability density functions (PDF) for ex-vessel calculations performed with containment codes. The sensitivity analysis was conducted for the original sample used in the UA, and for an increased sample size, to evaluate the consistency of the correlation coefficients and resulting PDFs. The results showed rather large standard deviations for some of the output parameters because of cliff edge phenomena in the material slumped to the lower plenum. Such results suggested the use of one-sided tolerance limits to set the initial conditions for posterior containment UA.

In 2019, the International Atomic Energy Agency (IAEA) launched the five-year Cooperative Research Project (CRP) I31033 to advance the understanding and characterization of sources of uncertainty and investigate their effects on the key figures-of-merit (FOMs) of the severe accident code predictions in water-cooled reactors (WCRs). Energy Software Ltd. (ENSO) contributed with an assessment of the uncertainty propagation in a long-term SBO scenario postulated for a CAREM-like integral PWR (iPWR). The study aimed at demonstrating the RELAP/SCDAPSIM/MOD3.5 (RS3.5) capability to carry out a BEPU calculation of a Severe Accident scenario in a single sequence from operational conditions to Reactor Pressure Vessel (RPV) creep rupture. The uncertainty analysis was conducted with the IUA2.0 module integrated into RS3.5 code, using the input-propagation methodology with a statistical description of the uncertainty proposed by Wilks. A group of 20 input parameters and 10 Figures of Merit (FOM) were selected for the assessment. The input parameters included boundary and initial conditions, material properties and code correlations, and the FOMs were related to the time of the main events and the fission product releases. To support the results, the Pearson, Spearman and Kendall correlation coefficients were analyzed for the selected input parameters and FOMs by using scalar values tables of the significance level. The relevant conclusions of the assessment are first, the importance of using the relative time for FOMs during core damage progression, and second, that including the Kendall formulation is advisable because it seems less dependent to singular data.

In an effort to address the inherent uncertainties in severe accident codes used in nuclear accident analysis, Energy Software S.L. (ENSO) has embarked on an innovative project aimed at developing a stepwise methodology for analyzing these uncertainties. The results obtained from two international projects, IAEA CRP I31033 and HORIZON-2020 MUSA, have provided ENSO with a solid foundation to identify and quantify the sources of uncertainty in severe accident analyses. However, they also revealed significant limitations, such as excessive computation time, multiple simulation errors, and truncation effects.

The proposed approach is stepwise, applying the Wilks/Wald method in two consecutive phases: the first linked to the "in-vessel" phase, where the accident prior to vessel failure is analyzed, and the second focused on the "ex-vessel" phase, examining subsequent events in the containment. With this methodology, ENSO plans to develop a model of a Generation II four-loop Westinghouse PWR reactor to simulate the "ex-vessel" phase of a low-pressure station blackout (SBO) accident using the MELCOR code. For the "in-vessel" phase, a previously created model with RELAP/SCDAPSIM will be used. The project also encompasses the development of pre- and post-processing tools with Python for uncertainty analysis with MELCOR, and the selection of input parameters based on probability distributions to apply the Wilks/Wald methodology.

Implementing the stepwise methodology will allow the identification and quantification of the sources of uncertainty at different stages of the accident, providing critical information for decision-making in emergency situations and for designing future research in this field.

In the process of severe accidents of the pressurized water reactor (PWR), the high temperature on the primary side of the steam generator and the high pressure difference between the primary and secondary sides can pose a high risk of creep rupture for the heat transfer tubes. Once the heat transfer tubes rupture, they lead to a bypass of the containment shell, causing radioactive materials to be directly released into the environment, and creating significant safety issues. The present study uses software to investigate the changes in parameters on the primary and secondary sides of the reactor under severe accident conditions, and calculates the rupture probability of the steam generator heat transfer tubes based on these parameter changes. MELCOR is used to simulate the station black out event (SBO), which has great impact on the heat transfer tubes. The main objective of the study is to clarify the changes in pressure and temperature on both sides of the heat transfer tubes under this accident condition, and to calculate the creep failure probability. The influence of microscopic and large cracks on the failure time of the heat transfer tubes is calculated. Considering that the surge line has the same failure risk, this study also shows the probability of the heat transfer tubes failing before the surge line.

For safety studies of nuclear reactor cores, a Finite Volume thermal-hydraulic code with pourous media approach named THYC was developed by EDF to simulate steady-state flows of a two-phase mixture within the core of a nuclear reactor. The steady-state solution is obtained through a fictitious transient. Despite a relatively low individual computational time, many statepoints are considered in the safety studies, which can represent significant CPU time.

To speed up the computation, one possible objective is to reduce the number of iterations (i.e., the number of time steps) required to reach convergence. In the present work, the idea is to train a neural network to predict steady-state solutions and use this prediction to initialize the transient computation. This method allows to combine the advantages of neural network prediction, in terms of rapidity, with that of the THYC model, in terms of the physical validation of the solution.

To evaluate the potential of this method, a simplified 1D code was designed. It simulates a two-phase water-vapor flow in a heated channel. A neural network was trained to predict the solution fields from imposed boundary conditions. In this paper, we give a presentation of the methodology, the database selection process, the structure of the neural network and the optimization of the network's hyperparameters. The highlight of this work is that, by introducing spatial frequencies in the error for the optimization of the neural network, we significantly reduce the number of iterations by 50% to 80%.

To establish an efficient and high-precision model calibration method for enabling virtual entities (e.g., digital twins) to synchronously track performance variations of physical counterparts, this study focuses on the feedwater heater model. A weighted method and the artificial bee colony (ABC) optimization algorithm are integrated to develop an intelligent calibration method, along with a prediction-correction-based acceleration technique. The proposed methods are tested on feedwater heater models from typical Gen II, Gen III, and Gen IV reactors. Results demonstrate that the acceleration technique improves calibration efficiency by 60- fold. The intelligent calibration method outperforms traditional manual approaches in both efficiency and accuracy: for single-load-case calibration, it achieves 99% accuracy within 0.5 seconds; for multi-load-case calibration, 98% accuracy is attained within 1 second, and 99% accuracy within 10 minutes. Thus, the calibration method and acceleration technique developed in this study meet the requirements of high efficiency and precision for virtual entities such as digital twins.

Printed Circuit Heat Exchangers (PCHEs) are one of the promising heat exchanger candidates for advanced nuclear reactors and high-temperature applications. This study focuses on developing machine learning models to predict the friction factor (f_D) and Nusselt number (Nu) for semi-circular zigzag channels in PCHEs. A number of data samples were collected from published literature including the channel geometric characteristics and operating conditions. Various machine learning techniques were employed, involving the linear regression with non-linear transformations, kernel methods (Kernel Ridge Regression and Support Vector Regression), and Artificial Neural Networks (ANNs). The Kernel Ridge Regression (KRR) model with a polynomial kernel of degree 6 achieved good performance for Nu prediction, with an R² score above 0.99 and low percentage errors (MAPE < 4%, MPE < 20%). This developed model can contribute to optimize the heat transfer performance of PCHEs, but its application is limited to helium as the working fluid. However, the ANN model for fD prediction, while showing promising results (R² > 0.97, MAPE < 10%), exhibited high maximum percentage errors (MPE > 100%), indicating challenges in predicting the friction factors (fD) less than 0.1. This study highlights the potential of utilizing machine learning models to improve the PCHE design. However, the expanded datasets covering a wider range of geometric configurations, working fluids, and operating conditions, and a detailed analysis of the input feature distribution would be useful to improve model accuracy.

The main drawback of empirical correlations in thermal-hydraulics system codes is that the prediction capability heavily relies on the quality of the data and vastness of the experimental data employed in the study. Therefore, in a long-term research program to improve the accuracy and reliability of the safety analysis methods of nuclear reactors at Hanoi University of Science and Technology, we have developed a method of integrating data-driven and machine learning models with a computing program on the basis of the following two basic modules: (1) experimental data analysis and predictive model development based on experimental data; (2) code develoment module based on conservation equations using the finite volume element method. In this paper, we introduce preliminary results with some case studies.

Thermal stratification in large enclosures is an integral phenomenon to nuclear reactor system safety. Currently, the effective model for thermal stratification utilizes a multi-scale method that integrates 1-D system-level and 3-D CFD code, which offers thermal stratification details while supplying system-level data across various domains. Nonetheless, harmonizing two codes that operate on different spatial and temporal scales presents a significant challenge, with high-resolution CFD simulations requiring substantial computational resources. This study introduced a data-driven coarse-grid turbulence model based on local flow characteristics at a significantly coarser scale targeting improved efficiency and accuracy in reactor safety analysis concerning thermal stratification. A machine learning framework has been introduced to expedite the RANS-solving process by coupling of OpenFOAM and TensorFlow, which entails training a deep neural network with fine-grid CFD-generated data to predict turbulent eddy viscosity. The feasibility of the developed data-driven turbulence model was proven through the SUPERCAVNA experimental facility problem validation.

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

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

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

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

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

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

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

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

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

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

The main outcomes of the following projects will be discussed:

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

Interface Capturing simulations are becoming a more practical tool for complex flow analysis due to significant improvement of flow solvers, pre- and post-processing tools as well as rapid development of high-performance computing capabilities. This creates exciting opportunities to study complex reactor thermal hydraulic phenomena. This presentation will focus on the history and review of numerical flow simulation approaches in recent years, capabilities development and validation as well as the applications to practical problems of interest. We will discuss typical computational methods used for those simulations, provide some examples of past work, as well as computational cost estimates and affordability of such simulations for research and industrial applications. New generation methodologies are required to take full advantage of those capabilities to greatly enhance the scientific understanding of complex flow phenomena in various conditions relevant to nuclear energy applications.

This study explores critical heat flux (CHF) in flow boiling on a downward-facing surface with non-uniform heat conductance properties. While prior investigations have examined non-uniform heat conductance surface, this research uniquely focuses on the downward-facing configuration, addressing an area that has been less explored. Additionally, the study expands the scope of surface modifications compared to earlier work, enabling a broader analysis of CHF behavior and non-unform heat conductance surface.

A significant advancement in this research is the introduction of a new CHF enhancement concept designed to prevent CHF initiation due to local hot spots. This concept is experimentally validated by measuring temperatures at both upstream and downstream regions of the heated surface, ensuring that local temperature variations and CHF dynamics are thoroughly understood. The methodology incorporates low thermal conductivity tape for surface modifications, chosen for its versatility and ease in creating various thermal conductance conditions.

Experimental results reveal a substantial CHF enhancement of up to 20%, highlighting the effectiveness of the proposed approach. These findings provide valuable insights into boiling heat transfer improvement strategies, particularly for downward-facing applications, and demonstrate the practicality of mitigating CHF triggers through innovative surface designs and precise thermal management.

Furthermore, the concept designed to prevent CHF initiation due to local hot spots can be applied to enhance the in-vessel retention capacity of a pressurized vessel. Since heat flux is non-uniform across different regions of the vessel, this approach could improve the overall CHF behavior, thus enhancing the thermal management and safety of pressurized vessels.

In this study, the effect of oxidation on the critical heat flux (CHF) of a vertical copper surface with four 16mm x 3mm x 2mm horizontal grooves and an bare copper surface will be investigated. During the experimental process, the copper surface will continuously oxidize with the repetition of experiments. For the bare copper surface, the oxidation rate is faster, as evidenced by its rapid darkening and loss of metallic luster, and the corresponding CHF value gradually increases with the surface oxidation. For the grooved copper surface, the oxidation rate of the surface is similar to that of the bare copper surface, but the oxidation rate within the grooves is much slower than that of the bare surface. The CHF value will remain stable within a certain range for a period of time before rapidly increasing to another level and continuing to remain stable. Overall, compared to the bare surface, the grooved copper surface has an enhancing effect on CHF, which is due to the change in macrostructure causing changes in bubble behavior, while oxidation also enhances the CHF of the copper surface, which is due to the change in the heating surface, thereby indirectly changing the bubble behavior.

Critical heat flux (CHF) prediction is essential for high heat systems like nuclear reactor and two-phase flow systems for enhancing dependability, safety and efficiency. CHF imposes design and operational restrictions due to safety concerns. However, there is work to be done to develop a reliable and effective CHF model. Machine learning techniques can find patterns and correlations in big datasets but lacks in explaining physical laws underlying CHF. Traditional ML operates as black-box which may result in physically unrealistic predictions, when applied to unforeseen circumstances and show instability. To address this, Physics-informed machine learning (PIML) integrates physical principles into PIML framework. While conventional ML uses only data, PIML integrates data-driven learning with knowledge from physical model. The goal of this study is to see the impacts of different physical models on reducing black-box nature of PIML and improving its interpretability for CHF prediction. In this study, four physical models for calculating CHF (Zuber, Katto-Ohno, Biasi and 2006 lookup table) were used as physical part of PIML and coupled with different pure MLs. A big amount of experimental data was used for the training and validation purpose of pure MLs and PIMLs. A thorough investigation has been carried out to assess (1) the predictive power of PIMLs and (2) compare the physical behaviors of mass flux, pressure, diameter, ratio of length-to-diameter and inlet subcooling on CHF from the PIMLs and pure MLs. This work shows the necessity of selecting a suitable physical model for approaching a robust and dependable PIML model.

Curved tubes such as U-bend, helical coil tubes etc., are commonly seen in the design of steam generators, heat exchangers in power plants due to their compactness and better heat transfer characteristics. Prediction of Critical heat flux (CHF) in the curved tubes is necessary in the system design for improved performance and safe operation. The occurrence of CHF is dictated by the geometric and operating conditions such as channel diameter, mass flux, subcooling, operating pressure etc. In the present study, a dimensionless number (De) determines the heat transfer and fluid flow characteristics in the curved geometries. The effect of De on the occurrence of the CHF in the curved tubes is given less attention in the literature. To this end, the simulations are performed using the two fluid model framework coupled with a wall heat flux partition (WHFP) model. The range of mass fluxes varies from 1500-3500 kgm^-2s^-1 and the degree of subcooling varies from 10-30 K . It was observed that the secondary flows created in the bent tube due to the centrifugal acceleration causes the fluid to undergo greater turbulence and flow separation in the bent region as De increases that enhances the heat transfer characteristics. A higher De causes the fluid velocity to increase which in turn causes the wall temperatures to drop. This further lowers the vapour volume fraction thus delaying the occurrence of CHF ensuring the safe operation.

Critical heat flux (CHF) is a primary power limiting criterion for water-cooled nuclear reactors, which must operate below the CHF conditions to allow sufficient thermal margin during normal operation. Accurate prediction of CHF is important for reactor design and safety analysis to determine safety margins under normal operating conditions, evaluate the maximum sheath temperatures of fuel bundles under anticipated operational occurrences, and predict the consequences under design basis accidents. To date, the most accurate CHF prediction method covering the widest range of flow conditions is the CHF look-up table. The latest version of the CHF look-up table was published in 2006. It was developed based on the world's largest CHF database as of 2005. Since then, a large number of CHF experimental studies have become available, allowing for the existing CHF prediction methods to be further improved. An upgraded CHF look-up table was derived from the expanded CHF databank containing 172 data sets with 42667 data points. The upgraded CHF look-up table provides good predictions for the range of flow conditions covering the normal operation, anticipated operational occurrences, and anticipated accident scenarios of water-cooled nuclear reactors. The upgraded CHF look-up table is expected to aid in improving subchannel and system thermalhydraulics analyses of reactor safety margins and consequences of postulated loss of coolant accidents for water-cooled reactors. The upgraded CHF look-up table also improves the accuracy of predictions under conditions related to fusion reactor divertor applications, and conditions corresponding to supercritical water reactor applications.

In the period of an intensifying energy crisis, the development of nuclear energy is of great importance. Fuel assemblies are critical components of reactor cores. As an innovative fuel type, Helical Cruciform Fuel (HCF) offers a larger heat transfer area per unit volume compared to traditional round fuel rods. Its unique helical structure enhances fluid flow and heat transfer capabilities. Additionally, the periodic contact formed by the helical structure provides self-supporting functionality, eliminating the need for position supporting and simplifying the structure of reactor core. These advantages make this novel fuel rod a focal area of research in Small Modular Reactor (SMR). In the course of extended operation of reactors, the pressure vessel may undergo deformation, causing displacement between the fuel rods and the pressure vessel and resulting in eccentricity. The heating curves of nuclear rod bundles typically exhibit non-uniform heating patterns in the reactor core. Unlike uniform heating methods, non-uniform heating introduces greater uncertainty in the location and values of critical heat flux (CHF). In this paper the RPI boiling model combined with the Eulerian-Eulerian two-fluid model are used to investigate the subcooled boiling and critical heat flux (CHF) heat transfer characteristics of HCF under non-uniform and uniform heating. Besides the influence mechanisms of different eccentricities on HCF with non-uniform and uniform heating power curves are explored. The findings of this paper will provide valuable insights for further research and practical applications of HCF.

As the third stage of the large break loss of coolant accident, whether the core can achieve effective cooling in the reflooding stage is the most important stage to prevent the large break accident from developing into a serious accident. The motion and heat transfer behavior of the droplets play an important role in the development of the re-submergence stage and are the key factors limiting the peak cladding temperature ( PCT ). The measurement of droplet parameters can provide necessary data support for the development of relevant mechanism models and the safety analysis of reactors under water loss accidents.
Conductivity probe is an effective means to measure droplet parameters. In this paper, the droplet characteristic signals and processing methods under different flow patterns are analyzed from the perspective of algorithm. The deviation of droplet measurement by three-sensor conductivity probe under different droplet size, droplet velocity and probe offset was studied by visualization technology. The relationship between uncertainty and different physical quantities was obtained. This investigation can provide a reference for droplet measurement using conductivity probes under re-submergence.

Accurately detecting coolant level and state is essential for ensuring nuclear reactor core safety, playing a critical role in early incident detection and severe accident prevention. However, level information alone is insufficient to fully evaluate the heat transfer performance and flow conditions of the coolant, necessitating additional void fraction measurement. To complement this, this study developed a horizontal pipe experimental system simulating air-water stratified flow and designed a fiber optic sensor-based device to measure the liquid fraction, and validated its concept through experiments. The device is integrated into the air-water stratified flow system, designed to minimize flow disturbances, and to detect the liquid level by utilizing differences in heat transfer characteristics between the media. This enables the calculation of local flow velocity and liquid fraction. Preliminary operation tests demonstrated stable performance under water flow rates of up to 5 L/min and air flow rates of up to 50 L/min. Experiments varying the power applied to the heating wire revealed distinct heat transfer characteristics, which were also observed under cooling conditions. Additionally, the sensor was able to measure interface movement in various flow environments, particularly confirming a tendency for the interface to be detected in regions with rapid temperature gradient changes. The system, leveraging the high spatial resolution of the fiber optic sensor, provides reliable data while validating the measurement method. Future research will construct a steam injection environment to enable phase detection and steam quality measurement in multiphase flow conditions similar to nuclear systems, further enhancing its practical applications.

In a reactivity-initiated accident, the reactor power might increase exponentially, following an escalation period. The larger is the insertion of reactivity, the shorter is the period.

Under such conditions, critical heat flux (CHF) limits cannot be described using models and correlations derived from and validated against steady-state experiments.

In this work, we present experimental results of transient CHF conducted on surfaces with different finish and wettability in subcooled (10, 50 and 75K) flow boiling conditions at atmospheric pressure.

The results confirm that, for slow transients, the transient CHF approaches the steady state value, which depends on surface finish. However, for fast transients, the CHF values are found to be independent of the surface finish and mostly increase with decreasing period. This observation suggests that the triggering mechanism of the boiling crisis in transient conditions may be different from the one under steady power inputs. It also undermines the rationale of models and correlations that aims at estimating the transient CHF on a certain surface starting from the steady-state CHF values.

Accurate measurement of wall shear stress and near-wall velocity profiles is critical for understanding the thermal and hydraulic performance of pressurized water reactor (PWR) fuel bundles. This study introduces an innovative experimental setup that employs Molecular Tagging Velocimetry (MTV) for direct measurement of flow velocity and gradients fields within a surrogate PWR bundle. The system integrates high-power optical fibers for laser light delivery and a borescopic imaging system embedded within the bundle rods, minimizing distortions and enabling local, high-resolution measurements.

Custom-designed optics ensure efficient laser coupling and delivery through optical fibers, achieving over 85% transmission efficiency. A custom borescopic system, paired with refractive index-matched (RIM) materials, minimizes imaging distortions caused by material interfaces. Preliminary results demonstrate the system’s capability to capture high-resolution flow patterns with a spatial resolution of approximately 10 m/pixel. A small-scale 3×3 rod bundle prototype with an instrumented central rod has been developed and tested under controlled flow conditions, validating the imaging and laser delivery systems.

This work lays the foundation for implementing MTV techniques to measure velocity gradients and wall shear stress in a realistic reactor geometry. By overcoming optical and spatial limitations, this setup provides a pathway for precise experimental data to support advanced numerical simulations. Future efforts will focus on deploying this methodology
in larger-scale and more complex PWR surrogate bundles.

The reliability of the primary cooling circuit in a pressurized water reactor (PWR) is crucial for safe operation. Debris fretting, caused by solid particles in the coolant, can damage fuel rods and lead to the leakage of fission products into the primary circuit coolant. Optimized filters in the fuel assembly bottom nozzle (BNO) can capture debris while minimizing pressure loss and reducing clogging risk.

To investigate the cooling flow through the bottom nozzle and filter, Magnetic Resonance Velocimetry (MRV) was employed using a 3 Tesla magnetic resonance imaging (MRI) scanner. The study focused on a simplified bottom nozzle, filter, and the leading edge of a 5x5 fuel rod bundle. MRV provided high-resolution measurements of 3D velocity vectors and 3D Reynolds stress tensors, without requiring optical access to the complex filter structure. The pressure drop across the filter was measured separately. Wire-like particles were introduced sequentially, enabling precise tracking of their positions and analysis of their impact on flow.

At a Reynolds number (Re) of 50,250, and with up to 100 particles, the filter test resembled standard conditions. A clogging scenario was created by introducing 240 additional particles at Re = 20,000. Using MRI data, the clogged filter’s geometry was reconstructed for CFD implementation. These CFD-grade measurements provide unique experimental data for validating particle motion and clogging models. Time-averaged velocity and Reynolds stress tensor data provide critical insights into how particles and agglomerations influence flow through the filter and around fuel rods, informing design improvements for enhanced reactor safety and efficiency.

Both direct simulation heater and indirect simulation heater rods have an extensive history of being used for a variety of nuclear reactor thermal-hydraulic testing including rod bundle CHF measurement, natural circulation cooling, reflood quenching analysis, flow induced vibration, and experimental observation of different thermal hydraulic phenomena. Despite this, recent studies show that there are a number of challenges associated with the use of indirect heaters, including the introduction of major measurement uncertainty and non-conservatism in CHF prediction, the non-prototypical and non-conservative peak cladding temperatures during blowdown, reflood quenching, and other transient heat transfer temperature measurements during both heat up and quenching processes. Most of these issues can be directly correlated to the physical composition of an indirect heater. Because the internal heating element is surrounded by highly conductive boron nitride or magnesium oxide, heat loss in the axial, lateral, or circumferential directions will be substantially large in case of any local heat transfer transient and/or deterioration events, causing the inaccuracies and non-conservatism observed in many experimental tests. This paper further details the challenges and non-conservatism of indirect simulation heaters, including experimental and simulation based examples. The non-conservative measurements were also verified with different modeling computations. Comparatively, the performance of direct simulation heaters is assessed in a similar manner, with results confirming that the use of indirect heaters poses a great risk to safety analysis and accurate thermal hydraulic analysis.

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

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

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

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

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

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

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

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

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

The summary and conclusions are presented below.

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

This study presents a computational fluid dynamics (CFD) methodology for simulating boiling at bubble scale without prescribed flowrate in the context of the quenching fabrication process. The approach is based on the Volume Of Fluid (VOF) method and the incorporation of phases sources allowing the water/steam phase change. This method is based on the use of an adaptive mesh allowing a very high refinement along the water/steam interface during the development of the bubble. The primary objective was to validate this methodology by comparing CFD results with experimental data in pool boiling cases. The quantity of interest is the wall temperature at different heat fluxes and for different surface topologies (roughness is modeled explicitly). Analysis shows promising agreement between CFD results and measurements regarding the wall temperature for low heat fluxes. This work demonstrates the model’s ability to create and develop multiple interacting bubbles as well as predict relevant wall temperatures for low heat fluxes.

This work represents a significant advancement towards developing a methodology to numerically assess the heat removal capability of a given surface as a function of its topology.

The strengthening of pool boiling heat transfer capacity is of great significance. As an active enhancement technology, ultrasound can effectively enhance the boiling process by influencing the growth and detachment of bubbles. With low cost and simple operation, it has a broad application potential while the mechanism still needs to be studied.

This study presents a multi-physical model which considers acoustics and fluid dynamics based on Multiphysics software. The volume force term is added to describe the nonlinear effects caused by the sound field; and the level-set method is used to track the phase interface that brings the saturated boiling bubble behavior under the influence of ultrasound. As a commonly used numerical value in the industry, the ultrasonic frequencies are set to 20,28 and 40 kHz. Numerical simulation has found that the acoustic streaming caused by ultrasound can cause the enhancement of fluid flow which generate shear forces on the bubbles. The acoustic streaming also make perturbations on the surface, which can accelerate bubble detachment and further enhancing the surface heat transfer capacity. As a result, the heat transfer efficiency has a considerable increase. Increased frequency and ultrasonic power can effectively enhance the acoustic streaming then act on the heat transfer process. Meanwhile, experimental research has been carried out to verify the results of numerical simulations. The derived conclusions could be useful for the application of ultrasonic treatment on boiling heat transfer.

Nucleate pool boiling is a useful heat transfer technique present in many engineering applications such as space and aircraft industries, thermal design of electronic components, refrigerants, and nuclear reactors. Despite its wide use, the physical mechanisms linking heat transfer to bubble dynamics and turbulence remain largely unexplored. Depending on the subcooling temperature, which is the difference between the wall temperature and the liquid saturation temperature, bubbles may either depart, merge rapidly, coalesce slowly or shrink. These complex dynamics affect heat transfer and lead to intricate vortex patterns in the flow, influencing velocity and temperature statistics.

The present work uses Direct Numerical Simulations with an in-house solver for incompressible multiphase flow to study the effects of subcooling on boiling behavior. The level set method captures the interface between liquid and vapor phases, while a computer vision algorithm was developed to track bubble properties. The findings show that subcooling temperature significantly impacts heat flux and bubble behavior, which in turn alters the flow's vorticity structures, producing ring vortices or irregular patterns. Statistical analysis is provided to better understand these complex interactions, shedding light on the relationship between bubble dynamics and heat transfer in boiling flows.

The authors are grateful for the financial support by the National Aeronautics and Space Administration (NASA) Grant number: 80NSSC21K0470 monitored by Dr. David F. Chao. We thank NASA Ames Research Center and NASA Advanced Supercomputing Division (NAS) for their generous allocation on Pleiades to perform three dimensional CFD simulations.

Predicting maximum sheath temperature accurately is important from the safety perspective (to demonstrate fuel and fuel channel integrity). The geometry and dryout mechanisms for CANDU fuel channels differ significantly from the light water fuel bundle assemblies. A new Eulerian multiphase computational fluid dynamics wall boiling model is being developed to model liquid film thickness-induced dryout in CANDU fuel channels. The operating conditions in CANDU fuel bundles are especially challenging because they span over a range of two-phase flow regimes. The proposed liquid film thickness-induced dryout model, which leverages advanced boiling closures for water at high pressures, was assessed in an earlier study using single-element CHF tests performed at Stern Laboratories, and encouraging results were obtained. This paper documents the model development for CANDU fuel bundles, its implementation in the STAR‑CCM+ software, and a qualitative assessment of the predicted dryout power using data from heated tests on the modified 37-element CANDU fuel bundle configuration. The approach adopted in the analyses is anticipated to yield advanced predictive capabilities that can be leveraged to improve traditional reactor safety analyses.

High resolution computational fluid dynamics (CFD) studies of turbulent flow boiling are challenging due to the simultaneous requirements that turbulence and bubbles are resolved adequately. These interface-capturing simulations can support the understanding of the mesoscale mechanisms of the heat removal process. Detailed analysis of such simulations has been demonstrated to uncover important physics of the heat transfer process and support macroscopic heat transfer model development. One area in which CFD can support the understanding of complex fluid mechanics is high heat flux boiling conditions. In pursuit of this goal, a benchmark problem has been developed to mimic the conditions of a previously completed high resolution experiment. The topic covered in this paper is a scoping study designed to assess the performance of the selected approach applied to this problem. The domain in this case is accurate to the experiment, with realistic Reynolds number, inlet subcooling, and heat flux. However, for simplicity and computational cost concerns, only a limited number of nucleation sites are considered. The mesh design is covered in detail to illustrate the consideration of both liquid turbulence and nucleate boiling. The bubble dynamics and associated wall temperature are compared against experimental values to assess the suitability of the approach for future multiple nucleation site studies and to which surface conditions the approach is generalizable. Based on the observed results, CHT & microlayer evaporation models are expected to introduce additional physics to improve predictions. Therefore implementation of these models is evaluated for future work.

This research aims at simulating and contributing to understand the generic physics which could occur in nuclear spent fuel pools during loss of cooling accidents. Based on the limitations inherent to the Direct Numerical Simulation approach in terms of Rayleigh number and geometry, we also intend to provide relevant reference results for RANS simulations.
This paper presents in a stepwise layout the different necessary components to simulate the natural convection of water in a cuboid domain combining vapor bubbles dynamic and evaporation through the free surface resulting into a descent of the water level.

Heat transfer due to evaporation is accounted for using the model presented by W. H. Hay et al. (2021), while the related mass transfer relies on a new remeshing procedure which attributes the descent proportionally to all cells by re-meshing the grid at each time-step. This allows to avoid any field changes at the boundaries whilst distributing the error along the height. This remeshing procedure, although apparently simple, involves a change in the temporal discretization of the governing equations. On the other hand, an Eulerian-Lagrangian approach is implemented and allows to compute the motion and growth/shrinkage of vapor bubbles while the effect of the bubbles on the fluid is accounted via momentum and energy exchanges between the two phases in a two-way coupling.

First, we detail the different models implemented. We then present a validation and verification procedure against analytical, experimental and numerical results. Finally, we present and discuss results of both models separately and combined.

This study presents a computational fluid dynamics (CFD) methodology for simulating multi-regime two-phase flow in an in-line tube bundle. The approach is based on a mixture multiphase model, incorporating scale separation between large interfaces resolved according to the mesh size and the dispersed phase modeled using a three-dimensional slip velocity to account for kinematic disequilibrium between phases. The primary objective was to validate this methodology by comparing CFD results with experimental data. The quantities of interest are the Power Spectrum Density of the forces applied in a tube and the local distribution of the void fraction around an instrumented tube. Analysis shows good agreement between CFD results and measurements regarding the quantities of interest investigated. This work demonstrates the consistency and the reliability of the methodology for two different mixtures: Water/Air and fluids simulating water/steam. In addition, this approach reduces the computational complexity in comparison to two-fluid models while maintaining a good accuracy in predicting two phase flow topology in the tube bundles.

This work represents a significant advancement towards developing a one-fluid formulation methodology for simulating complex multi-regime two-phase flows in tube bundles, with particular emphasis on studying fluid-structure interaction (FSI) in the U-bend region of steam generators.

The OpenSTREAM computational environment is a new open-source platform designed to facilitate efficient and collaborative development and validation of one-dimensional, multi-field, two-phase flow simulation models across research institutions. It includes several simulation frameworks: a mixture model, a two-fluid model, a three-field model, and an advanced four-field model of annular two-phase flow. The current implementation supports single-component, incompressible, steady-state, and transient boiling two-phase flows in single straight channels under reasonable simplifying assumptions. The two-fluid model solves a six-equation system governing mass, momentum, and energy conservation for each phase, capturing hydrodynamic and thermal non-equilibrium effects. The three-field model follows a classical framework (vapor, drops and film) for annular two-phase flow, while the advanced four-field model explicitly represents both the base liquid film and dispersed disturbance waves as separate fields. In all solvers, field interactions and wall closure models have been implemented either from well validated models from the literature or from simple considerations, providing a foundation for future collaborative improvements. Simulations of a representative boiling water two-phase flow case using all simulation frameworks show consistent and reasonable predictions. A comparison with the TRACE system code demonstrates that the implemented two-fluid solver produces reliable and consistent results. Finaly, the validation exercises from the original four-field model development are reproduced. OpenSTREAM, along with its validation and application database, will soon be publicly available on dedicated GitHub repositories under the permissive MIT license.

Since innovative reactors' flow conditions and geometries may differ from those of conventional reactors, the applicability of the models and correlations used in the design works should be appropriately checked. In the design phase, there is a high possibility that the flow conditions and geometries will be changed. Therefore, applying detailed numerical simulation is expected to achieve efficient design works. In nuclear reactors, two-phase flow will appear in many situations. Confirming the applicability of models and correlations for two-phase flow conditions is important. The detailed two-phase flow simulation codes must be useful to confirm the applicability of two-phase flow models and correlations. However, there is no established methodology to properly validate detailed two-phase flow simulations. We have, therefore, started the research project to develop a methodology for validating detailed two-phase flow simulation codes. In this project, we have developed two-phase flow measurement technologies to obtain detailed information on the gas-liquid interface and two-phase flow database by using developed measurement technologies and performing detailed two-phase flow simulations for the developed two-phase flow database. We will compare detailed two-phase flow simulation results and the two-phase flow database and discuss the proper methodology with the reactor manufacturer. Finally, we will investigate the validation process of a detailed two-phase flow simulation code. In this presentation, we will show the outline of this project and future plans.

Two-phase liquid-gas flow has a wide variety of applications in the nuclear industry, including active thermal control systems, steam generators, and nuclear reactors. In order to model and predict the pressure drop and flow regimes in a reactor core, the void fraction must be accurately predicted. This paper presents a new mathematical model that can accurately predict the two-phase void fraction requiring only knowledge of the geometry of the channel, liquid and vapor mass flow rates, and properties of the working fluid. The predicted void fraction is validated by void fraction data collected using an in-house capacitance sensor and a unique vertical, air-water flow calibration loop. Compared to measured void fraction data, the new mathematical model has a better performance than commonly used models such as Lockhart Martinelli model, Wheeler model, Chen model and homogeneous model.

Two-phase flows are critical in various industrial settings, including nuclear reactors, heat transfer systems, chemical processes, and wastewater treatment. Air bubbles within these flows enhance mixing, enable oxygen transfer, and alter heat fluxes. In nuclear reactors, bubble dynamics and oxygen transfer play pivotal roles in containment cooling, pressure control, gas stripping, hydrogen/oxygen management, corrosion control, and thermal-hydraulic modeling, making a comprehensive understanding essential. Computational Fluid Dynamics (CFD) simulations offer powerful insights into these systems beyond what sensors alone can achieve.

This study examines the impact of two diffuser configurations on flow mixing and oxygen transfer in a 1.3-meter water-filled reactor with 16 air diffusers. Two configurations were tested: all diffusers active (configuration A) and only the central lines active (configuration B), both operating at a flow rate of 20 m³/h. Simulations using OpenFOAM's twoPhaseEulerFoam solver incorporated the two-film resistance model with Clift’s mass transfer coefficient. Experimental data collected on void fraction, bubble velocities, and liquid flow supported validation.

Findings showed strong alignment between simulated and experimental results for void fraction and velocity profiles, allowing for detailed analysis of flow patterns. Configuration B demonstrated a 15% reduction in oxygen transfer efficiency experimentally, while CFD predicted a 24% decrease, effectively capturing the trend. These CFD simulations offer pre-construction insights into diffuser performance, informing design decisions on hydrodynamic interactions and oxygen transfer efficiency. Future work will enhance model accuracy and explore additional flow rates and dynamic aeration configurations.

While many advanced reactor-design concepts do not rely on subcooled boiling, the Pressurized Water Reactor still dominates world-wide installed capacity and accurate prediction of transient and steady characteristics of the nucleate boiling heat transfer regime has a first-order impact on the reactor design efficiency and safety margins. After more than 70 years of study there remain gaps in knowledge and uncertainties in empirical models and correlations. With the continued increase in available computational power, interface resolving high-fidelity simulations have become an important tool in closing these gaps in knowledge. Numerical investigations at practical scales involving thousands of bubbles are now possible. However, resolving the micro scale surface topology and roughness necessary for in situ prediction of bubble inception and inertial growth remains computationally out of reach for the foreseeable future. In this work, we will present a closure model of vapor bubble nucleation waiting time and inertial phase growth aimed at reducing uncertainty in existing high-fidelity numerical investigations of nucleate boiling heat transfer. The model is based on a simplified energy and force balance on the extant vapor bubble retained in the micro-cavity. The inception of vapor bubble growth considers local thermodynamic effects and surface conditions and is formulated as a renewal time. Care has been taken to provide a numerically stable and computationally efficient closure to the higher-order thermal hydraulic simulations.

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

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

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

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

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

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

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

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

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

In the frame of the French ANR Post-Fukushima ICE program (“Interaction between Corium and Water”), a series of Fuel Coolant Interaction (FCI)/Ex- vessel Steam Explosion (EVSE) integral tests have been performed at the KROTOS facility of the Severe Accident platform PLINIUS, located at CEA-Cadarache. Complementary, thermodynamic and thermophysical properties of prototypic corium chosen for integral tests KROTOS have been measured on VITI facility (CEA-Cadarache) and ATTILHA facility (CEA-Saclay). Post-test analyses on corium steam exploded debris have been performed through high-resolution X-ray diffraction measurements done on the MARS beamline at the SOLEIL synchrotron radiation source. The experimental research of ICE program was focused on fragmentation, dispersion of corium jets and formation of debris beds mechanisms, steam explosion energetics, corium oxidation and solidification mechanisms. The up-grade of KROTOS facility and new configurations for VITI and ATTHILA facilities to answer to the ICE program scientific objectives will be presented. Three integral KROTOS tests will be described and the knowledge gained for FCI/SE modelling will be discussed. A special focus will be done in the assessment of thermodynamic and thermophysical corium properties measurements and modelling. The very new results obtained concerning the corium final solid state and the cationic composition fluctuation that occurs in the U_1-xZr_xO_2-y solid solution will be presented.

Advanced Technology Fuels (ATF) cladding materials have become a key research focus worldwide, particularly at KIT, due to their promising potential to enhance reactor safety under accident conditions. FeCrAl and Cr-coated Zirconium alloys are designed to reduce hydrogen generation at least at the beginning of severe accidents (SAs). Their application in Small Modular Reactors (SMRs) is particularly relevant, as SMRs are emerging as a safer alternative to traditional Nuclear Power Plants (NPPs) due to their reduced core inventory and advanced safety systems. Within the framework of the EU SASPAM-SA project, this study focuses on the analysis of hypothetical SA scenarios involving a generic integral Pressurized Water Reactor (iPWR) with natural circulation, using the integral code ASTEC v3.1.2, ASNR all rights reserved, 2024. This code models thermohydraulic and physicochemical phenomena, allowing a detailed assessment of the accident progression from the initial event to the potential release of the radioactive material to the environment. This study specifically examines the performance of Zircaloy-4 and ATF cladding materials, focusing on their influence on the core degradation and hydrogen generation. The results show distinct hydrogen release kinetics for ATF materials compared to Zircaloy, emphasizing the impact of cladding properties on hydrogen production and safety margins during SAs.

The optimization of post-accident management in case of a severe accident (SA) is complex, particularly in what concerns handling of nuclear materials. After a SA, most of nuclear materials remain within the nuclear power plant (NPP) units in a solidified state, usually referred to as corium. The dismantling phase entails cutting such corium chunks into manageable pieces without causing an unnecessary radioactive remobilization to the gas phase that might result in further source term to the environment. This is currently the stage to be faced shortly in the Fukushima Daiichi site.

The OECD/FACE (Fukushima Daiichi Nuclear Power Station Accident Information Collection and Evaluation) project is dedicating significant resources to finding the best process for dismantling the site. Achieving this goal requires exploring different techniques and protection measures. This work is an exploratory analysis on how effectively an overlying water layer could absorb particulate material generated during laser cutting, in preparation for retrieving fuel debris from the affected units. Using the SPARC-Jet code, an in-house extension of SPARC-90 (Suppression Pool Aerosol Removal Code), the influence of uncertain boundary conditions on water retention efficiency has been studied. The focus was on factors such as carrier gas flow rate, particle size and concentration, pool temperature, and water depth. Preliminary results suggest that injection spot diameter and injected gas mass flow rate lead to higher Decontamination Factor (DF) values. However, from the cleaning efficiency standpoint, variables such as water temperature or depth should not be a concern, as their effect is very minor.

This work is being carried out in the context of severe accidents mitigation in sodium-cooled fast reactors (SFRs). The corium (set of molten core materials) formed may be transferred through discharge tubes to the lower part of the reactor vessel, towards a core-catcher. However, this corium could reach the core-catcher in the form of a hot jet (3000K), which could lead to local ablation of the core-catcher. This risk must therefore be taken into account to ensure that the core-catcher retains its integrity during corium relocation phase.

In the present work, the effect of the core-catcher geometry on its ablation process by a hot jet is investigated experimentally. Experiments were conducted on HAnSoLO setup, with simulating materials (transparent ice /jet of water). The experimental conditions were determined to be as representative as possible of those of a nuclear reactor. The geometric features of the core-catcher which are studied are its inclination and roughness. It has been observed that these two parameters significantly influence the ablation phenomenon, and in some cases can increase the ablation rate.

Sodium Cooled Fast Reactor (SFR) is one of the most promising Gen-IV concept for earliest deployment, owing to vast operating experience worldwide. Currently operating SFRs adapted partial core meltdown as a design basis for Core Cather (CC). However, the safety criteria for Gen-IV demands demonstration of safe mitigation of whole core accident and accordingly the CC design shall consider in-vessel retention and long-term cooling of the degraded core. Whole core retention would impose considerably higher heat flux and the CC need to withstand higher thermomechanical loads. To fulfil this requirement, development of an advanced core catcher has been taken up at IGCAR, India. The main objective is to develop and qualify a refractory protective layer for the CC, which is compatible with sodium and can withstand severe thermal transients expected during corium relocation.

Based on several tests in-house, refractory magnesia was identified as a candidate material for protective lining on a stainless-steel substrate. Dedicated experiments were conducted with magnesia test specimens to study i) long-term sodium compatibility, and ii) resistance to thermal shock under simulated accident conditions. Based on the microstructure and phase analysis, the sodium compatibility was assessed whereas degradation of the specimens was determined from the destructive/ non-destructive tests before and after the experiments. Results showed the magnesia specimens to have excellent sodium compatibility and good resistance to thermal shock, indicating the magnesia lined CC to be a potential option as advanced CC for future SFRs. Design concept, experimental methods and important results are discussed in the paper.

Based on both the internal-trigger and external-trigger experiments conducted by Shanghai University of Electric Power, air entrainment is proved to be a significant factor that affects the triggering on the surface of molten droplets during fuel-coolant interaction (FCI). In this study, based on the Rayleigh equation, the mass ratio of steam to entrained air, and the disturbance amplitude and interface pressure difference at the vapor-liquid interface under different working conditions are calculated. The relationship between the air volume, the disturbance amplitude and the trigger time of the molten droplet, and the relationship between the interface pressure difference and the trigger strength of the molten droplet surface are thus analyzed. It is revealed that the air entrainment can stir the disturbance amplitude, thereby reducing the trigger time of the molten droplet. The variation of the trigger strength of the molten droplet surface is consistent with that of the vapor-liquid interface pressure difference. In order to further exhibit and verify the phenomenon, the breakup of the steam envelope with air entrainment is simulated by Moving Particle Semi-Implicit (MPS) method, and it is quantitatively estimated that under the air mass ratio of 50% and 90%, the air entrainment may reduce the trigger time of the steam envelope by nearly 1.4ms.

Flashing induced instabilities in vertical systems are one of the most common phenomena that can take place under low-pressure and low-power conditions. Typically, the vaporization process is trigged in the adiabatic riser due to the drop in the hydrostatic pressure. The physics of the flashing induced oscillations have been widely studied experimentally and numerically. However, most of the studied systems have been operated under natural circulation conditions which imposed restrictions in isolating the effect of the flow velocity in the process. Hence, the effect of the flow velocity during forced convection remains uncharted.

In this work, we study flashing induced instabilities under low pressure and low heat flux in a vertical pipe. The tests are conducted in a test loop consisting of a single horizontal heated channel followed a 5 m vertical inverted U-tube section. Sinusoidal flashing induced instabilities have been detected as the flow transitioned from stable single-phase to stable two-phase state. At low power, the oscillations are triggered by flashing and enhanced by subcooled boiling. As the power increases, boiling and flashing coexist. The oscillations amplitude and characteristics as a function of applied power are presented and explained physically.

High-Temperature Gas-cooled Reactor (HTGR) has the advantages of inherent safety and supplying high-temperature process heat. Two-phase flow instability may occur within once-through steam generator (OTSG). Two-phase flow model plays a critical role in predicting the stability boundary. There are several two-phase flow models, including homogeneous equilibrium model (HEM), drift-flux model (DFM) and two fluid model (TFM), etc. DFM is much more precise than HEM when there is a significant velocity difference between the liquid and gas phases, while TFM is very complicated. Consequently, DFM is adopted to deal with the velocity of two-phase mixture region. According to the differences in friction factor and heat transfer factor, the convective heat transfer process in the secondary side of OTSG can be divided into three regions. These three regions are the subcooled water region, the two-phase mixture region and the superheated vapor region. A frequency-domain method is adopted for stability analysis. The essence of DFM is the derivation of void fraction from gas-phase mass conservation equations. The expression of void fraction subsequently leads to the formulation of expressions for mixture density, mixture mass flux and quality of two-phase mixture region and superheated boundary. The transfer function for pressure drop and inlet velocity is derived from momentum conservation equations of three regions using integral and small perturbation method. OTSG can operate stably at the designed power level. Compared to DFM, the stability boundary of superheated evaporation systems predicted by HEM is more conservative. The superheated evaporation systems become stable when distribution parameter or drift velocity increases.

Steam Generator Tube Rupture (SGTR) accident in lead-cooled fast reactor would result in high-pressure subcooled water jetting into the high-temperature melt pool in primary circuit. The intense phase change could trigger a steam explosion, seriously threatening the structural integrity within the reactor. However, there is still a lack of the theoretical study of key physical processes involved in the initial stage of the accident, limiting the development of the safety analysis programs. The study focuses on the three components which are water, steam, and liquid lead-bismuth and establishes theoretical models of jet breakup and phase change steam expansion. The characteristics of jet instability and the variation patterns of key parameters such as breakup time, breakup length and droplet diameter are analyzed. Additionally, the pressure impact on the melt pool generated by the phase change heat transfer and steam expansion is calculated. A comparison with existing research validates the model’s rationality. Results indicate that an increase in jet velocity reduces both the breakup length and droplet diameter, while jet radius has a limited effect on the characteristic parameters of jet. Further, the steam expansion causes the temperature drop and pressure summit in the melt pool. This research provides valuable guidance for assessments of the risk of steam explosion in SGTR accident and for the development of related safety analysis models.

This work describes a transient computational code for prediction of the heat transfer regimes and the channel wall temperature during a transient heating of the channel. The modelling includes the single phase and two-phase heat transfer regimes and post Critical Heat Flux film boiling calculations. The code is based on conservation equations and correlations from literature. A transient heating experiment of water flow inside a stainless-steel tube (1 m length and 8.3 mm in diameter) was used for validation of the model. The flow velocity inside the channel was about 3 m/s, the heating power was increased up to 38 kW and the exit pressure was almost atmospheric. During the experiment, the flow rate, the channel power and the local outer wall temperature of the channel were continuously measured. In the experiment, the channel was connected in parallel to a large bypass, and during the power increase, redistribution of the flow was obtained. Based on the continuously measured values, the model uses a suitable correlation for each regime to calculate the channel thermal parameters (coolant temperature and quality, and the wall temperature). In the single-phase regime, an over-prediction of the experimentally measured wall temperature was obtained, partially due to inaccurate temperature measurement. In the two-phase regime, a good agreement was obtained between the measured temperature values, the temperature trend in time and the model. A new correlation was proposed for the post-CHF regime, based on the calculated void fraction in that zone.

This article focuses on defining an axial void fraction from axial pressure profiles in two-phase critical flow discharges made with CriticAl Flow Test facility (CRAFTY), a novel separate-effect-test facility located in LUT University in Lappeenranta, Finland. Additionally, a method to approximate the flashing delay from the same axial pressure profiles is included. Axial temperature profile is used for detecting local superheat before flashing occurs along the axis of the discharge tube. The discharge tube in CRAFTY resembles the VVER-440 steam generator tube with inner diameter of 13 mm and has length-to-diameter (L/D) ratio of 350 for the used cases. A high subcooling case (ΔTsub~60 °C, pup ~ 8 MPa) and near saturation case (ΔTsub~5 °C, pup ~ 5 MPa) is used in the article. Both cases are rerun as well for increased certainty for analysis.

The axial pressure profiles can offer insights where the continuous liquid phase disperses into non-continous mist/droplet flow. At this transition zone lies the two-phase choke plane as the pressure information from downstream cannot travel into upstream anymore. In one-phase critical flow, the flow velocity reaches Mach 1 making the pressure signal stalled, unable to travel upstream. For two-phase critical flow this analogy is not correct. The two-phase sonic velocities are order of magnitude lower than either the liquid or gas phase.

The working fluid leakage accident in a supercritical carbon dioxide (SCO₂) Brayton cycle system of a reactor will result in the depressurization and discharge of SCO2 from the tube into the atmospheric environment, accompanied by critical flow phenomena of gas-liquid two-phase, which seriously threaten the heat transfer characteristics of the reactor core. The gas-liquid two-phase flow characteristics within the tube determine the size of the critical flow rate at the break. To accurately predict the critical flow rate, a testing technique must be developed to quantitatively measure the gas void fraction of the gas-liquid two-phase flow. In this paper, a phase state measurement method for gas-liquid two-phase flow during the depressurization and flash evaporation process of low conductivity SCO₂ is developed based on a wire mesh sensor assembly. Combined with the established SCO₂ depressurization and discharge experimental setup, the effects of factors such as temperature and pressure on the gas-liquid two-phase flow characteristics within the tube during the SCO₂ depressurization and flash evaporation process are explored. The distribution patterns of gas and liquid phases (bubble size, shape, and velocity, etc.) are analyzed to obtain typical flow patterns of the gas-liquid two-phase flow. Finally, based on the experimental results and key dimensionless numbers, a flow pattern map is plotted, and a prediction model for flow pattern transition boundaries is proposed. This lays the foundation for the study of critical flow phenomena.

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

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

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

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

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

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

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

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

In a specific model of nuclear power plant, the cooling water source for the Steam Generator Blowdown System's (TTB) regenerative heat exchanger is characterized by low undersaturation. This condition can easily lead to the occurrence of Two-Phase flow upon encountering disturbances. The resulting Two-Phase water hammer can cause vibrations that lead to pipe system failure and equipment damage. Severe vibrations induced by condensation-induced water hammer can occur in the return line of the cooling water to the deaerator inlet, potentially leading to conventional island faults. This paper first establishes a TTB system model using FLOMASTER software and conducts simulation calculations to identify the fluid medium conditions that are prone to condensation-induced water hammer. Subsequently, a mathematical model for condensation induction is established based on fluid characteristics, and local Two-Phase water hammer simulation analysis is conducted using FLUENT software. This study investigates the mechanism behind the original design scheme's condensation-induced water hammer and characterizes the pressure oscillation conditions and mass transfer characteristics associated with the destructive intensity of the water hammer. Sensitivity analyses are performed on parameters such as the pipe diameter, inlet flow velocity, temperature, and pressure of the cooling water pipeline. Finally, based on the aforementioned research findings, several feasible solutions are proposed to improve the original design and prevent the occurrence of Two-Phase water hammer. This study offers a universal approach to preliminarily assess the potential for condensation phase change water hammer in nuclear power plant fluid systems, enhancing the overall safety of the plant.

Accurate modelling of natural convection is essential for nuclear safety applications, particularly in passive cooling systems of Small Modular Reactors (SMRs). Reynolds-Averaged Navier- Stokes (RANS) models, widely used in industry, often fail to capture key buoyancy effects, limiting their accuracy. This study presents a new transport equation model for temperature variance dissipation, εθ, specifically designed for imposed temperature boundary conditions. The model aims to improve the estimation of εθ and the time scale ratio, which is crucial for thermal turbulence modelling. The proposed model is first evaluated a priori using Direct Numerical Simulation (DNS) data from Flageul et al. and then validated a posteriori for forced and mixed convection cases using DNS data from Abe et al. and Kasagi and Nishimura, respectively. Results show that the model provides a significantly better prediction of εθ near the wall and in the buffer layer, while also improving the estimation of temperature variance throughout the flow. In mixed convection, it effectively captures buoyancy effects and reduces errors compared to existing algebraic models.

Boiling heat transfer is an effective cooling technique that utilizes the high latent heat from phase change. The most critical parameter in boiling heat transfer is the Critical Heat Flux (CHF). CHF represents the maximum heat flux through nucleate boiling, at which a vapor film forms on the surface, impeding heat transfer and causing a rapid increase in surface temperature, potentially leading to surface damage. In previous studies, infrared thermometry technique was used to analyze the temperature distribution at the CHF on surfaces, measuring the 2D temperature distribution on surfaces. However, infrared thermometry technique has limitations: the substrate must be opaque to infrared, which restricts material choice, and the infrared camera must observe from below the substrate, complicating experimental setup and limiting use under nuclear reactor conditions. This study aims to overcome limitations of infrared visualization techniques by using fiber optic sensors to measure the 2D temperature distribution on surfaces. Fiber optic temperature sensors are easy to install, either by embedding them in surface grooves or inserting them into capillary tubes, can withstand high-temperatures of 600-700°C, and can measure thousands of temperature points simultaneously with high resolution (1mm between points, 100 Hz measurement speed), making them a promising alternative to infrared thermometry technique. In this study, high-resolution fiber optic sensors were used to measure the 2D temperature distribution on surfaces in real-time during boiling heat transfer. Measurements were conducted on micro-pillar surfaces and flat plates, with results showing a significant improvement in CHF on micro-pillar surfaces compared to flat plate.

Bubble growth and departure are critical phenomena in applications such as boiling heat transfer on heated surfaces and gas generation during water electrolysis. These processes play a significant role in engineering systems requiring efficient phase-change heat transfer and fluid management. While bubbles under static conditions have been widely studied, bubble behavior in dynamic environments involving acceleration remains less understood. Such conditions frequently occur in real-world scenarios, including mechanical vibrations in industrial equipment, seismic events, and rapid accelerations during emergency maneuvers. Understanding bubble growth and detachment under these circumstances is essential for improving system performance and ensuring reliability.

This study experimentally investigates bubble growth and departure on a plate subjected to controlled linear acceleration. High-speed imaging captures the dynamic evolution of bubbles from nucleation to detachment. The applied acceleration alters the force balance on the bubble, influencing its growth rate, interface deformation, and departure characteristics. Key parameters such as detachment radius and frequency are measured and analyzed under various acceleration conditions. The results show that acceleration-induced forces lead to asymmetric bubble shapes and earlier detachment, which could significantly impact heat transfer and fluid transport performance.

This research provides insights into bubble dynamics in accelerated environments, enhancing the understanding of multiphase flow behavior. The findings contribute to the development of more efficient and reliable thermal management systems in dynamic conditions, addressing critical challenges in engineering applications involving non-static settings.

The boron concentration stability of the Reactor Coolant System (RCS) is crucial for the safe operation of nuclear power plants. During the dilution operation, some equipment has a strong fluid retention effect on the replenished deionized water. This effect leads to the RCS boron concentration fluctuations, making it difficult to stabilize the RCS boron concentration. This paper utilizes a coupled model based on the lumped parameter method and computational fluid dynamics (CFD) method to study the dynamic response of the RCS boron concentration during the dilution process. The spatiotemporal distribution of boron concentration is first analyzed. Then, the effects of flow rate ratio, initial concentration difference, and initial temperature difference on the dynamic response of RCS boron concentration are analyzed. The results show that significant concentration stratification appears in the Volume Control Tank (VCT). A large amount of water is retained in VCT and then slowly released into RCS. Due to this phenomenon, the boron concentration of RCS takes a long time to reach the target value. Besides, the flow ratio and initial temperature difference should be increased to shorten the completion time in practice. The effect of the initial concentration difference is almost negligible.

Many applications involving in-line check valves in nuclear power plants require customized designs that deviate from standard offerings. These customizations require tradeoffs in performance, such as fast closure to mitigate water hammer effects, versus low flows to open the valve, reducing pressure losses. One lesser discussed phenomenon faced in the customization process is the reverse force on the valve obturator, pulling it into the flow direction, rather than the intuitive push in the direction of opening the valve. This force can be encountered when certain geometric and flow parameters coincide. To better understand the correlation, a select number of geometric parameters were varied using a sphere-in-pipe configuration, in a design-of-experiments (DOE) consisting of 81 configurations, intended to replicate the actuation forces on a check valve obturator, where the reverse force may be encountered. CFD simulations determined the force on the obturator, while maintaining a constant Reynolds number across the DOE cases. A mesh convergence study was performed, and steady state results were validated against transient results. The normalized force was mapped in response surfaces as a function of the normalized geometric parameters. Results showed a strong correlation between the obturator diameter and the orifice diameter creating zones of reverse force, suggesting what design configurations to avoid, while the correlation of the distance between the obturator and the orifice to the reverse force was weaker.

Accurate and efficient calculation of the fuel rod resonance cross section (XS) in the case of non-uniform temperature distribution is an important challenge in numerical reactor physics calculation when considering the neutronics and thermal hydraulics coupling. This paper studies the deep learning based global-local coupling resonance calculation method, which can accurately and efficiently calculate the effective self-shielding XS under the condition of non-uniform temperature distribution in the fuel rod. Through theoretical and data analysis, the input parameters and output parameters required for deep learning are obtained and the deep learning model is generated by training. The deep learning model is applied to the calculation of the local effective self-shielding XS in the global-local coupling resonance calculation method for considering the non-uniform temperature distribution in the fuel rod. Combined with the 2D/1D coupled transport method and the thermal hydraulics code, high-fidelity core neutronics and thermal hydraulics calculations are realized. The numerical results show that, compared with the calculation results directly using the ultra-fine group method, the calculation efficiency is increased by more than 20 times while maintaining the calculation accuracy.

The PAFS is passive cooling system to replace the conventional auxiliary feedwater and this system was adpoted as one of the cooling systems of the APR1000 and iSMR. This system removes decay heat from the reactor core by cooling down the secondary system of steam generator using a PCHX installed in the PCCT which has the role of ultimate heat sink. Due to the design characteristic, the cooling performance of PAFS is determined by boiling heat transfer outside PCHX and condensation heat transfer inside. In this paper, the prediction capability of SPACE and MARS-KS codes which currently used for thermal-hydraulic analysis in Korea were evaluated for cooling performance of PAFS. For this study, one of the IETs for PAFS performed by KAERI was selected. To evaluate the prediction capability of system codes, sensitivity calculations were performed using various boiling and condensation model options embedded within codes. As a result, it was confirmed that the system codes conservatively under-predict the heat transfer derived from the PAFS experiment, and it was concluded that further studies were needed to increase the accuracy in the future.

Boiling models are typically validated on greater than 5 mm hydraulic diameter channels as found in shell and tube boilers, or on microchannels with diameters less than 0.5 mm for CPU cooling. However, for the application of the Compact Steam Generator (CSG), which utilizes Printed Circuit Heat Exchanger (PCHE), the flow channels in PCHE are in the range of 1-2 millimeters, categorized as mini-channels. There is limited research on boiling in channels of this size. Due to the rapid advancement of simulation tools, particularly Computational Multiphase Fluid Dynamics (CMFD), the Volume of Fluid (VOF) interface tracking method has the potential to become a powerful tool for studying two-phase flow regimes in non-conventional flow diameters, such as mini-channels. In this study, both the experiment and simulation are performed and cross-validated against each other. R-134a is selected due to its similarity in density ratio to water under CSG conditions. A test loop has been built to conduct the experiment using a single mini-rectangular channel in a vertical configuration. The experiment measures pressure drop, and captures high-speed video of the flow patterns. The results show that the VOF simulation is capable of reproducing good bubble structures and void fraction distribution similar to those observed in the experiment. However, finer mesh sizes may be necessary to resolve small bubbles or droplets in flow regimes such as low-void bubbly flow or mist flow. The validated simulation results are then used to develop models applicable to boiling in mini-channels.

Small Lead-based cooled fast reactors have advantages of large natural circulation ability, great inherent safety and compact reactor structure, making it one of the main candidates of small modular reactors to produce electricity for communities or islands. However, most of present system analysis codes used for Small Lead-based cooled fast reactors are developed by modifying system analysis codes for pressurized water reactor or sodium cooled fast reactor. In order to improve the applicability and extensibility of system analysis code for small Lead-based cooled fast reactors, a multi-domain system analysis code is self-developed and validated in this paper. First, the physical models including the properties models, the flow and heat transfer models, power calculation models, steam generator models, main pump and so on are established. Then, a system analysis code named SASLFR aiming for the system performance analysis and safety evaluation of small Lead-based cooled fast reactors has been developed. Using the object-oriented programming language Modelica, SASLFR has characteristics of supporting modularized, multi-field physical model unified modeling, which greatly improves the system modeling efficiency, applicability and extensibility. Furthermore, SASLFR has been validated with variety of test problem covering fundamental models test and integration effect test. The simulation results agree well with the available analytical results and experimental data, which demonstrate that SASLFR has the capacity of simulating the basic physical process and complex system characteristics of small Lead-based cooled fast reactor. Hence, SASLFR can be used for the design optimization and safety evaluation of small liquid metal cooled fast reactors.

KAERI (Korea Atomic Energy Research Institute) has been operating an IET (Integal Effect Test) facility, which ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) with reference to the APR 1400 (Advanced Power Reactor 1400) for experiments for transient and DBAs (design basis accidents). An experiment for MSGTR (Multiple Steam Generator Tube Rupture) with failire of AFWS (Auxiliary FeedWater System) had been conducted at the ATLAS. The purpose of the experiment was to resolve a safety issue that multiple failure accident shus as loss of AFWS during MSGTR could lead to the damage of the core. Thus, the experiment aims at evaluating the importance of the PAFS(Passive Auxiliary Feedwater System) operation during postulated accident. The PAFS is adopted as one of the cooling systems i-SMR(innovative Small Modular Reactor) as well as passive cooling system to replace the conventional active AFWS. In this study, the experiment has been analyzed by Korea thermal hydraulic system analysis codes, which SPACE and MARS-KS the comparison with experimental and calculation results have been performed. In addition, the evaluation of PAFS cooling capability prediction will be disscussed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a MSGTR accident as well as the predict ability of the system analysis codes at the PAFS operation.

Passive safety systems, with their reduced dependence on operator action and external power supply, are considered to have higher reliability and safety than traditional active safety systems. However, concerns remain regarding whether these systems can adequately perform safety functions across diverse scenarios due to their lower driving force. Consequently, Korea has developed a robustness assessment methodology to identify potential degradation factors affecting the performance of passive safety systems and, ultimately, reactor accident mitigation characteristics. This methodology’s applicability was evaluated through the passive safety system of SMART100.

The developed methodology has not been validated against the experiments. Therefore, this study aims to validate the methodology through the PERSEO experiment, designed to test passive heat removal systems (PHRS). PERSEO was selected as a benchmark due to its application in OECD/NEA/CSNI/WGAMA’s international benchmark problem, which has validated its utility for such evaluations. Following the robustness assessment methodology, the experiment review includes a detailed examination of the geometry, operational conditions, experimental procedures, and thermal-hydraulic phenomena. For robustness assessment, the regulatory system analysis code of Korea, which is MARS-KS 2.0 was used, as it has demonstrated capability in simulating passive heat removal systems.

The conservative estimation model analysis predicted heat exchanger performance below experimental results, similar to previous studies. After modifications to the heat exchanger model, the best-estimation model more accurately reflected experimental heat removal rates. Moving forward, the optimized PERSEO model will facilitate evaluating various degradation scenarios, advancing experimental validation of the robustness assessment methodology.

As an effective heat transfer mechanism, nucleate boiling phenomenon has been widely studied in both experiments and numerical simulations. In water-cooled nuclear reactors, it is critical to obtain localized information of the surface temperature and heat flux at the fuel cladding, which requires high-fidelity numerical simulation. One of key issues for the high-fidelity simulation of nucleate boiling is microlayer model. Currently, many studies utilize the interface tracking technique to realistically simulate the effects of microlayer evaporation heat transfer to the growth of boiling bubbles. However, the approach requires very fine meshes and thus extremely high computational costs, which deter practical exploitation of the approach. In this study, to achieve the precise simulation of microlayer evaporation heat transfer with reduced computational cost comparable to conventional CFD, an implicit microlayer model along with a new conjugate heat transfer solution scheme is proposed for the region where the microlayer depletes. The transition region with the maximum thickness about 50 is treated by the mesh inflation for the near wall cell. To improve the simulation accuracy, the mech refinement is also adapted in the bubble interface. Obtained results show good agreements with experiment data including the coupling relation of wall temperature, wall heat flux, microlayer thickness and bubble geometry. However, this preliminary study is limited to a single bubble, which can be seen as a benchmark test. Later, further numerical simulations of multiple bubbles will be performed to implement the proposed model to the practically useable mechanistic wall boiling model.

During the operation of a nuclear power plant, fuel assemblies in the core need to be replaced periodically. In-containment storage racks in the refueling pool are installed for temporary storage of fuel assemblies during refueling shutdowns. It is important to ensure that the temperature of the water and the fuel cladding are within a reasonable range. Therefore, we study the flow and heat transfer characteristics in the in-containment storage racks under different operating conditions. In this study, the natural circulation phenomenon at the fuel storage racks in the containment is numerically investigated. The thermal-hydraulic characteristics of the refueling pool under different loading conditions are simulated based on the porous media method. The local peak temperatures of water and fuel cladding are obtained. The results show that the structure and arrangement of the in-containment fuel storage racks can satisfy the cooling requirements under various operating conditions. There is no boiling phenomenon occurring in the refueling pool. The local peak temperatures in the refueling pool are almost the same for both the full and single-load conditions. The results of this research can provide important guidance for optimizing the design of fuel storage racks in the subsequent work.

This study investigates the core inlet flow distribution in a 1/5 scale model of the APR1000 reactor using Computational Fluid Dynamics (CFD). The APR1000 is an advanced reactor design integrating proven and innovative technologies, developed for projects such as the Czech Republics's nuclear new-build initiative. The objective is to ensure uniform flow and pressure distributions across fuel assemblies, critical for maintaining thermal and mechanical integrity. A CFD model was developed using ANSYS CFX, incorporating a geometry and grid strucuture that replicates the reactor's key components. Porous media modeling was employed for complex regions like perforated plates of the core simulator, balancing computational efficiency with fidelity. Comparative analysis of CFD results, based on RANS-based Turbulence model, against experimental data revealed discrepancies in flow mixing phenomena, particularly in the core's outer regions. While the CFD simulations showed a 10% margin of error for overall flow distribution, limitations in capturing large eddy dynamics led to deviations in mixing performance. These findings emphasize the challenges of simulating turbulent flow in complex geometries. Based on these findings, the aplication of a Large Eddy Simulation (LES) model to the CFD analysis enabled a more realistic simulation of flow mixing characteristics. Compared to the results from the RANS-based turbulence models, the LES approach yielded analysis results that were more consistent with experimental data.

The safety relief valve is used in nuclear power plants to prevent the over-pressurization of the primary system. The valve is located at the top of the pressurizer, which can cause a temperature gradient between the top and bottom, which may cause different thermal expansion of the internal components of the valve. In this study, we intend to analytically compare the temperature differences due to changes in the surface heat transfer coefficient of the valve. The analysis was performed using ANSYS CFX, and the analysis was performed assuming three cases. The first case is when there is no heat loss in the valve body, the second case is when there is heat loss in the body, and the third case is when there is heat loss in the upper parts of the valve body. The CFD results demonstrate that insulation type and steam volume fraction significantly impact temperature distribution inside the valve. When heat transfer through the valve body was minimized, the temperature difference between the inlet and the upper valve region was reduced. However, in cases where heat loss occurred in the valve body or upper pilot valves, the temperature difference increased.

The hypervapotron is a water-cooled device extensively used in thermonuclear fusion reactors for managing ultra-high heat fluxes (20–30 MW/m²). Its effectiveness stems from boiling heat transfer mechanisms and fin structures that promote efficient thermal regulation. This study employs multiphase computational fluid dynamics (CFD) simulations using OpenFOAM to investigate the effects of geometric variations such as fin height and side slot on cooling performance and internal flow dynamics in hypervapotron cooling channels.

Results revealed that side slots significantly enhance thermal management by enabling bidirectional fluid ingress, maintaining consistent liquid inflow into fin slots, and optimizing bubble removal. The side slots further contributed to the formation of diagonal flow patterns, which improved heat dissipation even under intense heat flux conditions. Conversely, channels without side slots exhibited higher wall temperatures and reduced cooling performance due to limited liquid ingress.

Adjustments in fin height were found to critically influence the cooling performance. Taller fins improved vortex formation and enhanced liquid inflow, resulting in superior heat exchange. In contrast, shorter fins led to incomplete fluid exchange and vapor accumulation within the fin slots, reducing efficiency.

This work bridges the relationship between hypervapotron design parameters and thermal performance. It provides actionable insights for optimizing the cooling channels to manage extreme thermal loads in fusion reactor divertors and similar applications. By integrating these findings, future designs can achieve enhanced thermal management and structural integrity under high-heat-flux environments, contributing to the advancement of hypervapotron technologies.

The natural circulation flap valve is a critical structural component of the poot-type low-temperature heating reactor. Under accident conditions, the two-phase natural circulation formed between the system and the reactor pool effectively cools the core to achieve safe shutdown. The flow space enclosed by the core riser, natural circulation flap valve, and top space of the reactor core forms a transverse T-shaped tube configuration. The characteristics of internal bubble-phase separation are intrinsically linked to the natural circulation capability, which plays a pivotal role in reactor safety analysis. Therefore, this study conducts an in-depth investigation on the characteristics of bubbly flow separation in the transverse T-shaped tube using numerical simulation methods, exploring the effects of branch height, main/branch tube diameter, void fraction, liquid flow rate, etc. on the separation characteristics. The results indicate that there is significant phase separation behavior in the transverse T-shaped tube. Specifically, the void fraction of the branch decreases with the increase of the branch height, while the stable rate of the branch pressure drop fluctuation increases with the increase of the branch height. These findings provide critical support for understanding and analyzing the characteristics of the two-phase natural circulation through the valve under accident conditions.

The integration of several codes has been recently developed to analyze multi-physical and multi-dimensional phenomena in a nuclear power plant. SPACE, CUPID, and MASTER codes are coupled. SPACE code is a one-dimensional system analysis code. CUPID code is a component scale three-dimensional thermal hydraulic analysis code. And MASTER is a three-dimensional neutronic code. SPACE and CUPID have similar governing equations for thermal-hydraulics. Thus, SPACE and CUPID are coupled to formulate a single pressure matrix for the coupled domain. This coupling scheme is well-validated with various theoretical problems and separate effect tests. CUPID and MASTER codes are coupled with sharing major variables, such as heat flux and temperature. To verify this multi-dimensional and multi-physic coupled code at a plant level, the main steam line break accident in the APR1400-kind plant is analyzed using this couple code. The core and connected parts of hot-leg and cold-leg are modeled with CUPID and the neutronic part is modeled with MASTER. The rest of the parts including the secondary loop are modeled with SPACE. In addition, the same scenario is modeled with only SPACE code. Comparing transient results of the coupled code and SPACE code, it is found that the newly developed coupled code is well verified for plant-level transient.

The potential of small modular reactors to solve the increasing international need for carbon-free electricity can only be realized with the concurrence of national nuclear regulators, such as the United States Nuclear Regulatory Commission (NRC). A major component of NRC approval of each unique design is the evaluation model (EM) development and assessment process, which produces an EM to demonstrate that nuclear safety is maintained within design basis accidents. The model is supported by experimental data, which may necessitate the creation of new test facilities or the design and acquisition of new loops and equipment capable of increasing the state of knowledge for impactful phenomena. The need for new data is determined by the phenomena identification and ranking table (PIRT) committee and depending on the extent of experimental needs for unique design elements, may include integral effects test and/or separate effects test facilities to be constructed for the verification and validation effort. Once the need for supplementary experimental data is decided, cost and schedule often necessitate scaling of test loops. A rigorous scaling effort is then employed to determine the most important physical parameters and dimensions that best represent the targeted prototype facility, while ensuring the key phenomenon of interest are preserved. These test facilities may also require improvements to existing infrastructure which could introduce new challenges while expanding future capabilities.

The McMaster Nuclear Reactor (MNR) is a research reactor located in Hamilton on the McMaster campus. The main purpose is to supply the medical industry with isotopes for a wide range of applications, most importantly for cancer treatments. In order to ensure the safety and performance of the MNR it is crucial to provide a solid database for a new thermal-hydraulic Safe Operation Envelope (SOE). With the McMaster hydraulic loop, there is a perfect tool to support the new SOE. The aim of the current work is threefold: First, gain a better understanding of the thermal-hydraulic characteristics of MNR fuel assemblies. Second, validate and refine computational tools such as system codes and CFD. Third, develop an artificial flow resistance that mimics the pressure drop of actual fuel assemblies, enabling cost-effective scaling of experiments without the use of original fuel elements consists the third goal. To complete these tasks, a single assembly is placed in a closed water circuit and subjected to varying mass flow rates. Parameters such as pressure drop, pressure distribution, temperature, mass flow, and density are precisely measured. The insights gained from this research contribute significantly to the optimization of reactor design and safety analyses, ultimately enhancing the accuracy of thermohydraulic prediction models for nuclear reactors. This comprehensive approach bridges the gap between theoretical models and practical applications, advancing the field of nuclear engineering. The results of different setups are compared and discussed in this work.

Recently, in the thermal hydraulic safety research area of nuclear power plant in Korea, the safety analysis methodology development with improvement of safety analysis code, SPACE, is now promoting to establish an IBLOCA, which has a smaller break size compared to large break loss of coolant accident (LBLOCA), as design basis accidents. In order to apply the SPACE code to the system analysis on the IBLOCA transients, appropriate SPACE code improvement along with the development of a new safety analysis methodology is necessary. And, of course, improved SPACE code should be evaluated and verified to confirm its capability.

In this study, an integral effect test database was established by utilizing ATLAS test facility which was constructed and operated by KAERI to verify the improved SPACE code. Three kinds of IBLOCA was simulated under the operating conditions of OPR1000 nuclear power plant. During the transient simulation, the system showed very general thermal hydraulic behavior that can occur in the IBLOCA transient, including a loop seal clearing phenomenon.

From the test results, the major thermal hydraulic phenomena were investigated and evaluated for the system cooling capability with an operation of safety systems. In addition, the difference of system behavior during the transient simulation according to the different break simulation will be investigated. The present test data can be utilized to verify and evaluate the improved SPACE code for application to an IBLOCA, as originally intended for the purpose of this study.

Major challenge for lead cooled fast reactors (LFRs) is the performance of structural materials. Specifically, flow-accelerated corrosion and erosion (FACE) phenomena may lead to deterioration of reactor internal structures. A dedicated facility, the Separate Effect Test Facility for Flow-Accelerated Corrosion and Erosion (SEFACE) is under design at KTH Royal Institute of Technology, to obtain experimental data in liquid lead conditions with high relative velocities and high temperatures. The facility is supposed to provide data for a wide range of flow and thermal conditions to study material degradation over months long periods of time in autonomous operation. A reliable system therefore is needed to maintain and control the required liquid metal temperatures at various flow velocity conditions. The flow velocity conditions are achieved in SEFACE by rotating disks accommodating experimental specimens in the main cylindrical vessel. The thermal effect of the disk rotation is the heat dissipation in the liquid lead. To keep the required temperature conditions, simultaneous active heating and cooling system is needed. To minimize vessel penetration and interference with vessel internal structures, a heating cooling jacket design is being developed. The active heating and cooling are achieved by heaters and water cooling tubes attached to thin fins of the jacket. Various configurations of fins, heaters, and water cooling tubes and material choices are discussed with numerical simulations and modular experimental tests. The optimal design is then selected and used for the temperature control system of the SEFACE facility.

The OECD/NEA ATLAS (Phase 3) project, spanning from 2021 to 2024, is a collaborative international project focused on addressing thermal-hydraulic safety and accident management challenges associated with water reactors utilizing the ATLAS test facility.
As part of this project, the C2.3 test was conducted to simulate a steam line break (SLB) scenario, incorporating the operation of a passive auxiliary feedwater system (PAFS). The primary goal was to analyze thermal-hydraulic transients within the reactor coolant system (RCS) during an SLB event with PAFS activation. Key parameters such as system pressures, temperatures, collapsed water levels, and natural circulation flows were measured and analyzed. The test outcomes highlighted the effectiveness of the PAFS operational system in cooling down the RCS during an SLB. Additionally, asymmetric thermal-hydraulic behavior was observed under asymmetric cooling conditions induced by the SLB. Effective cooling of the primary system was achieved during the operation of the main steam safety valves (MSSVs) of Steam Generator 2. Following PAFS actuation, a continuous decrease in core temperature and primary pressure was noted. This study further involved comparing the simulation results of MARS-KS with the OECD/NEA ATLAS C2.3 test data to evaluate the predictive capabilities of MARS-KS.

For the licensing and safety criteria of nuclear power plants, the standards for nuclear fuel are undergoing changes due to the adoption of DEC (Design Extension Conditions) and safety concerns related to high burn-up fuel. As a result, multi-physics coupled safety analysis has become a significant issue with the introduction of new LOCA (Loss Of Coolant Accident) criteria. In this study, the behavior of fuel cladding under simulated LOCA conditions was investigated using a thermo-mechanical and thermal–hydraulic coupled experimental facility known as ICARUS. The oxidized cladding was utilized to study high burn-up fuel, with an oxidized cladding being produced through an oxidation process. To assess the impact of oxidation, two reflood experiments were conducted—one with fresh cladding and the other with oxidized cladding—and the results were compared.

In this study, a safety analysis on a Pressurizer Level Control System Malfunction (PLCSMF) in the Advanced Power Reactor 1000 (APR1000) was conducted to investigate the effects of the pressurizer spray and heater operation. The Safety and Performance Analysis CodE for nuclear power plants version 3.3 (SPACE 3.3) was used for the transient calculation of the PLCSMF event. Based on the APR1000 design data, a nodalization of the nuclear steam supply system (NSSS) was modeled, and the steady-state calculations conducted for the combinations of the initial operating conditions, such as the RCS pressure, RCS temperature, and RCS flow rate. The PLCSMF event with 12 sets of the initial operating condition combinations were simulated as assuming the maximum charging pump flow rate and the minimum letdown flow rate with several conservative assumptions. In addition, the effects of pressurizer spray and heater operation were analyzed for each PLCSMF event, assuming both operation and non-operation of these systems. The pressurizer spray increased the RCS peak pressure and delayed its reaching time, resulting in a conservative outcome. The pressurizer heater had a negligible effect on the RCS peak pressure, but accelerated its reaching time. In conclusion, the present study highlights the importance of assuming pressurizer spray operation in the safety analysis of the ARP1000 PLCSMF event is necessary to obtain conservative. On the other hand, all APR1000 PLCSMF events analyzed in this study revealed that the RCS peak pressure were below the acceptance criteria.

In a severe accident in a nuclear power plant, hydrogen is generated within the reactor core and released into the containment building. It could accumulate and mix with air, posing a risk of explosion with low ignition energy. Passive autocatalytic recombiners (PARs) aim to reduce the hydrogen concentration inside the containment building below the flammable limits to prevent such explosions. The hydrogen-air mixture enters the bottom of the PAR by buoyancy. It moves upward between vertical plates coated with platinum (Pt) or palladium (Pd), where hydrogen recombines with oxygen through a catalytic reaction. The exothermic reaction and heating of product gases add additional driving force to the natural convection, and the resulting steam-air mixture is discharged at the top. In this study, we modeled the heat and mass transfer, fluid flow, and chemical reactions between two vertical catalytic plates using computational fluid dynamics (CFD) simulations. While most previous PAR studies applied a constant inlet velocity condition, we introduced natural convection conditions at the inlet to investigate how the flow rate of the hydrogen-air mixture entering the PAR correlates with the hydrogen concentration. As the inlet hydrogen concentration varied from 1 to 6 vol.%, the flow velocity entering the PAR reached a maximum at 4 vol.% and showed a decreasing trend beyond this concentration. We analyzed these phenomena by considering the buoyancy of the hydrogen-air mixture, the natural convection resulting from the exothermic chemical reaction, and the molar reduction due to the generation of steam.

By eliminating the graphite sleeve, a dual-separate direct cooling of fuel compacts is achieved, significantly improving both fuel cooling efficiency and the thermal output of high temperature gas-cooled reactors (HTGRs). In this sleeveless fuel design, however, fuel spacers hold crucial aspects in supporting the fuel compacts and maintaining open flow channels during normal operation in a helium coolant environment. The design of these fuel spacers has been an urgent issue in the related research field. To minimize pressure drops in the helium coolant flow, this study introduces a ring type fuel spacers that stabilizes each fuel compact from four directions. Real-scale prototypes of the fuel components were fabricated using 3D printing. To examine the flow characteristics around the fuel spacers, pressure drops measurements were conducted in an inert gas flow. As computational fluid dynamics (CFD) simulations of the spacers, this work performed CFD modeling using the commercial software, Star CCM+. Coupled with experimental results, this work has developed a validated fluid model around the spacers. This CFD model is a key tool for future V&V studies in thermal-hydraulics analysis of the reactor core.

Molten salt reactors (MSRs) are emerging as a next-generation nuclear technology due to their high operating temperatures and enhanced safety. The high melting point of molten salt leads to rapid solidification upon release from the reactor, which helps limit its spread. However, additional safety measures are necessary to minimize the potential dispersion of molten salt, as its spread and solidification significantly influence the release of radioactive materials. Argonne National Laboratory (ANL) previously used the MELTSPREAD code, originally developed for corium behavior analysis, to simulate the spread of molten salt on flat surfaces, using FLiNaK as the working fluid. However, MELTSPREAD tends to underestimate the spread radius because it does not consider for remelting. In this study, computational fluid dynamics (CFD) is used to improve the simulation accuracy of molten salt spread and solidification by incorporating remelting. NaCl-KCl-UCl3, in which 10 kg is poured for 1500 s, was chosen as the working fluid, and the simulations utilized models including the VOF multiphase model, Solidification and Melting model, RANS model, and DO radiation model. The sensitivity of these models to various parameters affecting molten salt behavior was evaluated. The simulation results showed an increase in both the spread radius and heat transfer over time, with the leading edge of the molten salt solidifying first, which limited further spreading. To enhance the accuracy of molten salt leakage assessments, model selection and parameters such as initial temperature and flow rate were crucial. Future research will explore the inclusion of decay heat models.

Heat pipe reactors are pivotal for next-generation nuclear technologies due to their inherent safety and flexibility. However, temperature redistribution under operational variations poses challenges to system safety and efficiency. Existing data-driven methods for temperature prediction often neglect physical principles, raising concerns about reliability. This study proposes a hybrid data-physics framework integrating neural networks with heat transfer physics to predict peak fuel rod temperatures in heat pipe reactors. A 1/6 centre-symmetric component area was analyzed using Computational Fluid Dynamics simulations to generate temperature datasets. The framework combines a backpropagation neural network (data-driven) for rapid feature-point prediction with a physics-driven Virtual Circle Model. Validation against Fluent simulations demonstrated maximum reconstruction errors of 3.8 K (0.4% relative error) under normal conditions. A novel diagnostic protocol for single heat pipe failure localization was developed, leveraging measuring through-holes temperature differences (ΔT) to identify failure modes. The method also characterized dual-failure temperature escalation patterns, revealing peak fuel temperatures exceeding 1000 K in severe cases. This hybrid approach enhances interpretability, accuracy, and reliability in thermal monitoring, offering a robust tool for safety evaluation and design optimization in advanced nuclear systems.

The increasing demand for safer and more efficient nuclear power has driven the development of innovative small modular reactors (i-SMRs) incorporating fully passive safety systems. These systems eliminate the need for active safety components, ensuring reactor safety under accident conditions without external power or operator intervention. This study presents the i-SMR’s passive safety system, which integrates the Passive Emergency Core Cooling System (PECCS), Passive Auxiliary Feedwater System (PAFS), Passive Containment Cooling System (PCCS), and Containment Isolation System (CIS). These systems leverage natural forces such as gravity and thermal differentials to sustain core cooling and maintain containment integrity. The PECCS ensures core cooling through a depressurization and recirculation mechanism, while the PAFS facilitates long-term decay heat removal. The PCCS supports containment cooling using a dry cooling mechanism, and the CIS prevents radioactive material release by automatically isolating the containment vessel during accidents. The i-SMR’s passive safety design enables extended accident mitigation, maintaining core cooling and structural integrity for at least 72 hours without operator action. Compared to conventional large reactors, the i-SMR achieves a significantly lower core damage frequency (CDF), demonstrating enhanced safety and reliability. This research provides insights into system functionality, regulatory considerations, and the next steps for validation and deployment of this advanced passive safety technology.

Severe Accident simulation codes like ASTEC (Accident Source Term Evaluation Code) are essential for predicting nuclear reactor behavior under Severe Accident conditions. However, their high computational demands, particularly in thermal-hydraulic simulations (e.g., the CESAR module in ASTEC), can constrain their effectiveness. A significant portion of CESAR’s computational load arises from solving non-linear partial differential equations at each timestep using the Newton-Raphson iteration method. While Newton-Raphson convergence depends on having initial guesses close to the final solution, the current ASTEC implementation relies solely on values from previous converged states, without predictive insights.

This paper introduces a hybrid approach aimed at enhancing CESAR’s Newton-Raphson solver through machine learning (ML) models, which offer more refined initial guesses. Drawing on recent advancements, this approach explores using ML-based surrogate models to learn the intricate, non-linear relationships within transient conditions, aiming to reduce the number of iterations needed for convergence and potentially allow longer timestep intervals without sacrificing accuracy. The choice of surrogate model remains adaptable, seeking to balance predictive accuracy and computational efficiency within CESAR’s frequent initialization routines.

Preliminary results show that an ML-augmented approach can significantly reduce ASTEC's computation time without altering the convergence criteria, suggesting promising implications for broader applications in thermal-hydraulic simulations in nuclear safety assessments. Future work may also involve developing an ensemble of surrogate models, complemented by a classifier that dynamically selects the most suitable initialization based on the reactor state and accident phase, optimizing the solver’s performance across varying scenarios.

This study investigates advanced Physics-Informed Neural Networks (PINNs)—a-PINN, n- PINN, and can-PINN—applied to benchmark fluid dynamics problems, including lid-driven cavity, Taylor-Green vortex, and natural convection in a square cavity. Transfer learning and targeted modifications to governing equations significantly enhance PINN accuracy and stability at higher Reynolds and Rayleigh numbers. Despite computational complexity, results highlight PINNs’ flexibility and their potential as effective alternatives to traditional CFD methods.

This study proposes a new method for predicting Critical Heat Flux (CHF) and its position in fuel rod bundles, enhancing the safety analysis of heat transfer systems. By leveraging the attention mechanism of Transformers, CHF prediction is achieved using sequential features from continuous physical fields, improving accuracy in complex rod bundle subchannels while increasing model interpretability. This approach identifies key physical quantities and control volumes that influence CHF. Initially, typical CHF experimental data were collected, including inlet conditions, CHF locations, and corresponding CHF values for individual fuel rod bundles. Then, numerical studies were used to generate upstream physical field data such as flow velocity, pressure, and qualities, which are critical for CHF predictions. These quantities were organized sequentially along the flow direction and served as inputs for the Transformer. The attention mechanism processed these sequential features to achieve high-precision CHF predictions. Finally, interpretability analysis evaluated feature importance in the model's outputs, revealing the contributions of different physical quantities and control volumes. Results demonstrated that the Transformer-based model significantly improved CHF prediction accuracy within rod bundle subchannels. The interpretability analysis identified the most influential physical factors, validating the model's reliability and providing insights into the CHF occurrence mechanism. In conclusion, the Transformer-based attention mechanism enhances CHF prediction accuracy in reactor rod bundles and identifies critical influencing factors through interpretability analysis. These findings advance the understanding of CHF behavior under complex multi-subchannel conditions in rod bundles with Mixing Vane Grids(MVGs) and provide a valuable tool for safety analysis and thermal system optimization.

Heat pipes have long been recognized as an efficient and versatile way of transferring heat, outperforming traditional mechanisms. In recent years, the interest towards this technology has risen, driven by its potential applications in emerging technologies and energy systems. One notable field driving the implementation of Liquid Metal Heat Pipes (LMHP) is their application in nuclear reactors for space exploration. Due to their versatility, simple mechanisms, and reliance on physical processes that function independently of gravity, heat pipes are expected to play a crucial role in the future of nuclear reactors for space missions. In this context, accurately predicting the behavior of HPs under a broad range of conditions becomes essential. The frozen startup of LMHPs, in particular, poses a significant challenge due to the variety of interconnected physical phenomena involved.

Among the different modeling techniques present in literature, conduction-based models represent a relatively simple yet effective method of capturing the transient behavior of LMHPs.

This study examines the conduction-based model proposed by Yoo et al., with particular focus on the ability of the latter to reproduce the temperature evolution over the heat pipe during the frozen startup phase. The model has been validated against experiments conducted by Ponnappan and by Faghri et al. to ensure that the results align with those proposed by the original authors. Additionally, the model has been validated over novel experiments performed at MISOH1 facility at the University of Michigan.

Thereafter, the impact of selecting appropriate correlations for the material properties is systematically analyzed.

Under the US High-Performance Research Reactor (USHPRR) project, new fuel has been developed for several high-performance research reactors, in order to transition the reactors from highly enriched uranium (HEU) to low-enriched uranium (LEU). Fuel plates utilized in some of these reactors have a thin cylindrical shape containing uranium-molybdenum alloy fuel core in aluminum alloy cladding. The changes in the design of the LEU fuel element (e.g., thinner fuel plates, different coolant channel gaps) compared to the HEU fuel element necessitate the evaluation of the hydraulic performance of the LEU fuel plates.

Previously developed analytical models by Donald R. Miller and Wade R. Marcum both predict a critical flow velocity at which a cylindrical laminate plate would collapse, but the data available for validation is sparse. The flow velocity in all USHPRR fuel elements is significantly lower than the calculated critical velocity. This research, in addition to supporting the displacement sensor testing for the USHPRR hydraulic performance evaluation, aims at providing high fidelity data on the relationship between the maximum deformation at the leading edge of a plate, the pressure differential between channels of a single plate, and the critical flow velocity. The outcome of this research can help to further validate the existing analytical models

The emergence of Small Modular Reactors (SMR) has focused attention on small nuclear power plant systems. Minimizing the heat exchanger volume is crucial for developing such systems. The Printed Circuit Heat Exchanger (PCHE) is a key candidate for achieving this goal.

PCHE is manufactured through multiple processes. First, micro-channels are created on metal plates using chemical etching. These plates are then stacked and joined through diffusion bonding, forming a single module. With its micro-channel structure, PCHE achieves high heat transfer efficiency per unit volume and exhibits excellent mechanical strength.

The thermal-hydraulic characteristics of PCHE vary depending on the shape of the micro-channels. For example, zigzag channels induce high pressure drops due to their complex geometric structure but enhance fluid mixing, leading to superior heat transfer performance. Existing studies have primarily focused on analyzing turbulent flow conditions using supercritical CO₂ (SCO₂) and helium as working fluids, while studies on laminar and transitional flow regimes using water remain limited. Moreover, previous research has mainly emphasized heat transfer performance and pressure drop, lacking detailed flow pattern analysis.

This study focuses on the flow pattern analysis of zigzag channels. Flow visualization experiments were conducted under various Reynolds number conditions using Particle Image Velocimetry (PIV), and the results were compared and validated with Computational Fluid Dynamics (CFD) simulations. Through this approach, changes in flow patterns due to variations in the bending angle were analyzed, and a correlation for the friction factor, a key parameter in PCHE design, was derived.

One of the defining features of integrated energy systems is multi-energy coordination, where renewable energy sources, baseload energy sources, and storage technologies are flexibly combined to overcome the limitations of single energy systems. This highlights the importance of studying the characteristics and reliability of hybrid energy systems, for delivering stable and efficient energy. In this paper, we propose and analyze an innovative hybrid energy system consisting of the small modular reactor, concentrated solar power and a packed-bed thermal energy storage, with a closed Brayton cycle as its energy conversion system. The system's configuration and operational control strategies are detailed. A one-dimensional dynamic model of the entire system, as well as its main equipment, has been proposed to facilitate performance evaluation. Through simulations and analyses, we explore the behavior of key parameters under varying operating conditions. Simulation results demonstrate that the proposed hybrid energy system effectively meets critical performance requirements. First, the system exhibits excellent power flexibility, dynamically adjusting output power to match varying energy demands by operational control strategies. Second, it achieves strong reliability, maintaining stable operation under fluctuating conditions such as varying solar irradiance. The study proves the viability of nuclear-solar-thermal storage hybrid systems as a promising solution for achieving stable and reliable energy supply. It provides a foundation for understanding the dynamic behavior of such systems and offer guidance for their application in diverse energy scenarios.

The core interfacial friction model plays a dominant role in the prediction of core cooling and maximum clad temperature in PWR and BWR accident sequences. Due to a lack of precise void fraction data in such complex geometry, the code models of core interfacial friction still have a rather high uncertainty band. The FONESYS network of system code developers initiated an activity to compare the models of the ATHLET, CATHARE, LOCUST, RELAP5 and SPACE system codes. A first comparison of models is made in the domain of low flowrate in conditions where wall friction is small and void fraction depends only on a balance between buoyancy force and interfacial friction. Only the pre-CHF bubbly-slug-churn and annular flow regimes without drop entrainment are considered in the domain of void fraction (0 < α < 0.8). The effects of pressure (0.1 MPa < P < 12 MPa) and hydraulic diameter are investigated. The paper presents first the origin of the models. The results show that the various codes agree on the qualitative pressure effects with some differences on the hydraulic diameter effect.
However, significant differences are found in the very low-pressure domain. Possible effects of the wall heat flux are discussed. Further work is planned including validation and possible model improvements.

The LWR thermalhydraulic behavior in accidental transients encounters many types of bifurcating events (BE) with cliff-edge effects where some phenomena disappear and other appear. They are due to automatic control or operator actions such as SCRAM, ECCS actuation, pump start or stop, valve opening, etc, or to transitions between different flow regimes or heat transfer regimes. Such BE have a major impact on the main parameters of interest such as primary (and secondary) pressure and fluid mass inventory, and on the figures of merit such as a peak clad temperature. The prediction of all BEs and of the right occurrence time of BEs is the main challenge of experimental and numerical simulation tools. In the Phenomena Identification and Ranking Table (PIRT), the successive time phases of a transient or Phenomenological Windows (PhW) are first identified and they can be defined and bounded by some BE. In the scaling analysis performed to design Integral Tests Facilities (ITFs) and Integral Effect Tests (IETs), most scaling methods use dimensional analysis of scaling equations (mass, momentum, energy) at system scale with acceptance criteria on the ratio of non-dimensional numbers at reactor and experimental scales. This forgets the dominant role of BEs. One may use a mature system code to perform a more detailed scaling analysis of a transient and to focus on the respect of occurrence and timing of occurrence of major BEs as acceptance criteria for the design of an ITF and of IETs. Examples are given on some PWR LOCA analyses.

FONESYS is an international network of system code developers created in 2010 to share information on R&D, to benchmark codes, to discuss the Validation and Verification as well as the code scalability and uncertainty quantification. APROS, ARIANT, ATHLET, CATHARE, CATHENA, COSINE, MARS, LOCUST, MARS-KS, RELAP5, RELAP5-3D, SPACE, TRACE are the codes that were involved in the FONESYS activities: updating the state of the art, identifying issues, discussing envisaged solutions, sharing experience in 3-field models, transport of interfacial area, numerical issues and well-posedness, code uncertainty evaluation. Code benchmarking were performed on boiling channel with CHF and Post-dryout, two-phase critical flow, flow regime transitions in horizontal flow, core interfacial friction, core 3D-mixing effects and two-phase singular pressure losses. Many code improvements were implemented in the various codes following the benchmark activities. When the need of new experimental data was identified, FONESYS discussed with SILENCE experimentalists to define requirements of new instrumentation and new experiments. Future activities will focus on code scalability, core 3D modelling and validation, two-phase pressure losses, use of system codes for scaling analysis and applications to passive systems, SMRs and AMR. The present paper presents the major achievements and the motivations for the future activities.

The system codes can model 3D flow in a core either with 3D solvers in porous medium, or with sub-channel models, or even using cross-flow junctions between parallel 1D models. Such tools rise many questions on the modelling of mass momentum and energy transfers including diffusion and dispersion processes. The extrapolation of many closure laws from 1D to 3D flow and the formulation of wall friction and interfacial friction when the flow is not parallel to fuels rods require some attention. The FONESYS network of system code developers initiated an activity to compare the 3D-core models of the ATHLET, CATHARE, COSINE, LOCUST, RELAP-3D and SPACE codes in very simple situations. One considers first two adjacent fuel rod assemblies with different power and one calculates the vapor flow above a swell level at two pressures (1 and 7 MPa) to obtain either friction-driven crossflows or buoyancy driven crossflows (chimney effect). Other calculations include the two-phase region, a swell level and a dry zone above, the assemblies being connected to an upper plenum and a lower plenum, allowing different inlet flowrates as in a reactor situation. The impact of axial friction pressure losses and spacer-grid form losses on radial crossflows is shown. The homogenization of void fraction below a swell level seems very efficient. The sensitivity on radial pressure losses in non-axial flow is not very high but the uncertainty on the radial wall friction and interfacial friction is very high. From the analysis of these results, further investigations are planned.

The current system codes use 2-fluid models or 3-field models and predict wall friction and singular pressure losses by models developed for a unique mixture momentum equation. Two-phase multipliers exist that can extrapolate 2-phase pressure losses from 1-phase models but the repartition between phases is not modeled although it may play a very significant role on the result. This may result in a rather high uncertainty of predictions not only in high velocity flow but also in low flow situations encountered in natural circulation. The FONESYS network of system code developers initiated an activity to compare the predictions in a few basic singular geometries for which one can simply evaluate the single-phase pressure loss coefficient. In a reactor circuit, there are several locations with an abrupt area restriction or an abrupt area enlargement or even some plates behaving as a diaphragm. Some code prediction comparisons in such basic geometries are presented and analyzed in both vertical upward and vertical downward flow and in five flow regimes: 1-phase liquid, two-phase bubbly flow, low velocity and medium velocity slug-churn flow and high velocity annular-mist flow. The calculations were made with five codes: ATHLET, CATHARE, MARS-KS, RELAP5 and RELAP5-3D. Some differences are found between codes. The effects of nodalization are investigated and the impact on void fraction perturbation are analyzed. First conclusions are drawn on the reliability of predictions and some requirements for future well-instrumented pressure loss experiments are defined.

The Atomic Energy and Alternative Energies Commission (CEA) is a French Research and Development institution that plays a major role in the French nuclear energy program. CEA has been developing the CATHARE code for 45 years. It is an extensively validated thermal-hydraulic system code based on the 2-fluid 6-equation model. The CEA is currently developing new tools in CATHARE to facilitate its use for scaling analysis, which contribute significantly in the identification of dominant phenomena occurring during reactor accidental transients and in the design of experiments able to simulate major phenomena with minimal distortions.

Previous published work on scaling analysis with the CATHARE code focused on the primary mass and primary pressure equations to identify the dominant terms controlling the mass inventory and the system pressure. These terms were calculated “by hand” during the post-processing phase. A new tool enables this analysis to be carried out automatically during a calculation, by fetching the thermal-hydraulics quantities at code execution. It relies on a Python supervisor, based on the ICoCo (Interface for Code Coupling) standard, which extracts the required fields at each calculation time step. The fields are then manipulated using the MEDCoupling open source library. The integrated momentum equation along cooling loops is also added to the analysis.

This work describes the way this tool is applied to a SB-LOCA transient using the equations of mass, momentum and pressure. Figures of merit and available post-processing are presented, which enables the dominant phenomena to be quantified. Leads for future developments are given.

The models in system codes for LWR core interfacial friction still have a rather high uncertainty in some conditions due to a lack of void fraction data in such complex geometry. Usually, only axial pressure differences are available to test the interfacial friction. The FONESYS network of system code developers initiated an activity to compare the models of several system codes and found significant differences at very low pressure. Recently a new void fraction measurement technique based on the wire mesh sensor was implemented by CRIEPI in a 5x5 rod bundle and produced real void data. There are steady state data at 1, 3 and 7.2 MPa, with a wall heat flux from 5 to 45 kW/m² and a mass flux in the range 90 to 400 kg/m²/s. CRIEPI also performed boil-off tests with an outlet atmospheric pressure and heat fluxes in the range 2.4 to 7.2 kW/m². EDF and CEA performed validations calculations with the CATHARE code on these data. Several sensitivity tests are presented to investigate some possible reasons of the code-to data differences. Previous validations on data in boil-up tests and on air-water non-heated rod bundles data are used to investigate the various possible effects of fluid properties, of pressure, and of the wall heat flux. Preliminary conclusions are drawn on the validity and limitations of the current interfacial friction model and on possible ways to improve it.

In order to enhance the safety of nuclear power plants, it is required to establish emergency cooling methods for reactor accidents. In PWR, the using In Vessel Retention (IVR) method is considered to prevent melt-through in meltdown. In the IVR method, the cavity surrounding the reactor pressure vessel is filled with water in the IVR method to enable cooling by boiling heat transfer. The maximum cooling capacity of boiling heat transfer is determined by the critical heat flux (CHF), and improving CHF is crucial to implement the IVR. In this study, we found that forming a porous copper structure on the boiling surface improved the CHF to approximately 5 MW/m², which is about four times higher than that of an uncoated plain surface. The boiling experiments were conducted using a heat transfer surface with a diameter of 10 mm and porous copper structures with thicknesses from 0.5 mm to 3.4 mm under saturated temperature conditions at atmospheric pressure. It was observed that CHF increased as the thickness of the porous structure increased up to 2 mm, but decreased when the thickness reached 3.4 mm. In this presentation, we will discuss the factors contributing to the CHF improvement.

In the face of the escalating energy crisis, the advancement of nuclear energy assumes paramount significance. Fuel assemblies, crucial elements of reactor cores, play a pivotal role in this domain. During the long-term operation of reactors, deformation of the pressure vessel may occur, leading to displacement between the fuel rods and the pressure vessel, resulting in eccentricity. Studying the subcooled boiling and critical heat flux (CHF) phenomena within annular pipes and exploring the effects of eccentricity are crucial for clarifying the two-phase boiling processes in fuel assemblies and enhancing reactor safety. In the reactor core, usually the heating curves of rod bundles are non-uniform heating. Comparing to the uniform heating methods, the non-uniform heating mothods bring more uncertainness in CHF locations and values. Using CFD analysis software, in this paper the RPI boiling model combined with the Eulerian-Eulerian two-fluid model are employed to analyze the subcooled boiling and CHF heat transfer characteristics of annular pipes under non-unifrom and uniform heating. On this basis, the mechanisms underlying the effects of different eccentricities on annular pipes with non-uniform and uniform heating are explored. The findings of this paper provide valuable insights for a deeper understanding of heat transfer phenomena within fuel assemblies and offer guidance for their practical application.

During the operation of nuclear reactors, corrosion particles deposited on fuel cladding alter its surface characteristics, thereby influencing the critical heat flux (CHF). The modification of surface properties, such as roughness, wettability, and porosity, plays a significant role in determining the heat transfer efficiency and safety margins of the reactor. However, the absence of detailed structural parameters for the deposition layer has hindered a comprehensive theoretical analysis of the mechanism by which the deposition layer affects CHF. To address this gap, the present study conducted systematic pool boiling deposition experiments under varying deposition times and heat fluxes. The experimental results demonstrated that the average CHF of smooth rods was 1240 kW/m², with a deviation of less than 10% from model predictions, thereby confirming the measurement accuracy and stability of the experimental apparatus. This validation establishes a robust and reliable experimental platform for subsequent research on CHF behavior under deposition conditions. Notably, the average CHF of rods with deposition reached 1479 kW/m², representing a 19.2% enhancement compared to smooth rods. This significant improvement suggests that, in terms of CHF enhancement for fuel rods, corrosion particle deposition exerts a positive influence. These findings contribute to a deeper understanding of the complex interactions between surface properties and CHF, offering critical insights into the role of deposition layers in enhancing thermal performance.

Enhancing our understanding of the link between the near-wall void-fraction profile and nucleation characteristics at the boiling surface in subcooled-flow conditions is crucial for improving subcooled-flow and DNB models. This interaction is being investigated at a flow-boiling facility at MIT, which features a deionized water loop capable of operating at pressures up to 10 bars and mass fluxes up to 2000 kg/m²/s. The test section includes a rectangular flow channel (3 cm x 1 cm).

Subcooled nucleate boiling was generated using a custom-made heater, consisting of a sapphire substrate coated with a thin layer of chromium, housed in a heating cartridge mounted on one of the test section walls. An optical probe, translated with sub-millimeter accuracy to and from the heated surface, measured the void-fraction profile near the surface. Infrared measurements through the back of a sapphire substrate allowed for the quantification of wall temperature, heat flux, and relevant nucleation properties at several heat fluxes.

The collected measurements are the first step towards forming a comprehensive picture to elucidate the mechanisms linking the void fraction distribution in the vicinity of the boiling surface with the boiling process and the boiling crisis. Insights gained from this study may inform the development and validation of next-generation models for flow boiling simulations.

To prevent reactor vessel failure, the in-vessel corium retention through the external reactor vessel cooling (IVR-ERVC) has been adoptedas the severe accident mitigation strategy in nuclear reactors. In existing large nuclear reactor types, IVR-ERVC is conducted in a natural flow condition, which forms between the insulator and the reactor cavity. However, in Small Modular Reactors (SMRs) currently under development in the Republic of Korea, there is no insulator due to the integral design, and thus IVR-ERVC is conducted in a pool condition. Additionally, since the reactor lower head outer wall of SMRs is made of stainless steel, it is essential to study the Critical Heat Flux (CHF) phenomena occurring on stainless steel surfaces to ensure the safety and effectiveness of the IVR-ERVC process in these reactors. In this study, focusing on the integral SMR design, the experiment measured CHF under various conditions, including the effects of heater surface inclination and material properties for stainless steel. Experiments using an SS304 heater in pool boiling conditions are conducted to develop a modified CHF correlation, reflecting the specific characteristics of integral SMR, challenging existing models and contributing to safer nuclear power technology.

The critical heat flux (CHF) is a key thermal limit in water-cooled nuclear reactors and accurate and reliable modelling of boiling and CHF remains an unresolved challenge in nuclear thermal hydraulics. In large majority, CHF is still estimated using empirical models derived from expensive, full-scale experiments. Due to the empirical nature of the models, significant engineering margins are applied, restricting reactors to operate at a power level that is below their theoretical potential.

In the last few decades, computational fluid dynamics (CFD) models based on the multifluid Eulerian-Eulerian method and the heat flux partitioning approach have shown promise in reducing conservative design margins through more accurate predictions. However, the large number of modelling closures required (e.g., nucleation site density, bubble departure diameter) and the overfitting of the numerous constants on limited datasets has so far prevented developing a universally accepted, best-possible model and deliver the anticipated improvements.

In the last few years, advancements in measuring techniques have made possible detailed, small-scale measurements that enable the validation of boiling models at a level of detail that was not possible before. In this work, we have developed and implemented in MATLAB a heat flux partitioning framework and, leveraging these new data, assessed the most frequently used and recent heat flux partitioning models in pool and flow boiling conditions. Strengths and weaknesses of each model, and some physical inconsistencies are identified. Impact of uncertainty in closure models is quantified, improvements implemented and validated and areas for future development suggested.

Valve leakage failures are a common equipment problem in nuclear power plants. Such failures not only lead to economic losses, but also cause radioactive leaks in serious cases, resulting in major safety hazards. Currently, experimental data on valve leakage is very scarce, and the diagnosis of many types of valves is very complex, so it is necessary to supplement experimental data on valve leakage and select the best diagnostic algorithm.

This paper describes a novel experimental system designed to simulate Reactor Coolant Capacity Control System (RCV) upper loading line valve leakage faults, to verify the reasonableness of the experimental setup by analysing the experimental data and to explore the impact of leakage on the system performance. Within the realm of algorithmic analysis, this study evaluates the diagnostic efficacy of various algorithms, including logistic regression, random forest, and support vector machine (SVM) models. The empirical findings of the study reveal that the Random Forest algorithm exhibits the most superior diagnostic precision, achieving a remarkable accuracy of 99.82% in the detection of valve leakage incidents. Algorithms such as Support Vector Machines (SVMs) and Simple Bayes have unsatisfactory performance metrics, with diagnostic accuracies not exceeding the 95% threshold. Therefore, the latter algorithms are considered less suitable for diagnosing valve leakage faults.

This research not only enriches the experimental data but also offers a valuable reference for the selection of appropriate diagnostic algorithms. The in-depth investigation of valve leakage faults can serve as a robust safeguard for the secure operation of nuclear power plants.

One of state-of-the-art pressurized light-water small modular reactors, an innovative small modular reactor (i-SMR), is being developed in Republic of Korea. A steel containment vessel (CV) is adopted not only to prevent release of radioactivity material but also to reduce pressure and temperature of the reactor module. As a newly suggested passive safety system (PSS), a passive emergency core cooling system (PECCS) prevents water level reduction of reactor coolant system (RCS) using natural circulation without additional injection of coolant. The emergency depressurization valve (EDV) and emergency recirculation valve (ERV) which are installed on the wall of reactor vessel (RV) play as the natural circulation flow paths between RV and CV. The pressurized steam from the RV through the EDV is condensed in the CV by heat transfer on heat exchanger of passive containment cooling system (PCCS). The condensed water recirculates to the RV through the ERV. The level of condensed water is important physical variable because the difference of water levels between CV and RV determines recirculation flow rate.
Integral effect test facility is under basic design to conduct thermal-hydraulic validation tests to support the process of standard design approval (SDA) for i-SMR. The basic design of PCCS heat exchanger considering condensation heat transfer was completed and the scaling distortion in integral effect test facility against the prototype was evaluated. Four trains of PCCS heat exchanger with thermocouples to measure fluid and outer wall temperatures will be installed in the integral effect test facility.

The integrated full natural circulation SMR has a high degree of integration, high intrinsic safety, flexible arrangement and can be applied in various scenarios. However, integrated full natural circulation SMR also incorporates some innovative designs, such as the elimination of the Main Circulation Pumps (MCPs), the adoption of helical-coiled tube heat exchanger, the application of the passive safety design, and so on.

Due to these new characteristics of the integrated full natural circulation SMR, the existing system analysis code cannot be directly applied to the integrated full natural circulation SMRs, and the corresponding system analysis code still needs to be developed. RELAP5 code is a widely used system analysis code internationally and has been successfully applied to some SMRs. The RELAP5 code has been improved or added the relevant models that are required for the analysis of integrated full natural circulation SMR by authors.

The improved RELAP5 code is validated by an Integral Effect Test conducted by SNERDI. The comparison results of natural circulation tests at various power levels are shown in this paper. The results show that the improved RELAP5 code compares well with the test results, with a temperature difference of about 5 degrees. Additional test cases will be performed and further validation and evaluation will be conducted in the future.

The recent revision of Emergency Core Cooling System (ECCS) acceptance criteria, which incorporates Design Extension Conditions (DECs) and addresses high-burnup fuel safety concerns, has intensified the need for more accurate loss-of-coolant accident (LOCA) analyses. Previously, thermo-mechanical and thermal-hydraulic behaviors were evaluated separately, resulting in conservative estimates that limited insights into actual coupled phenomena. Implementing an integrated multi-physics approach now enables simultaneous characterization of these behaviors, leading to more realistic analyses and enhanced safety margins. In response to this need, the Korea Atomic Energy Research Institute (KAERI) developed the ICARUS facility to simulate fuel cladding behavior from the post-blowdown stage through the reflood phase of a large-break LOCA (LBLOCA). A Zircaloy-4 cladding and heater assembly, combined with controlled boundary conditions, replicates the thermo-mechanical and thermal-hydraulic environment of a reflood scenario. Real-time measurements of cladding surface temperature, deformation, subchannel fluid temperature, and water level are carried out. By varying heater power, internal cladding pressure, and the reflood initiation time, this study systematically evaluates the coupled phenomena, thereby offering critical insights into the multi-physics behavior of nuclear fuel cladding under LBLOCA conditions. By integrating thermo-mechanical and thermal-hydraulic analyses, this work moves beyond conservative assumptions and provides a more realistic understanding of cladding behavior under accident conditions.

The Molten Salt Reactor Experiment (MSRE) was a key nuclear project in the 1960s that demonstrated the viability of molten salt as a coolant and fuel. In molten salt reactors, understanding complex thermohydraulic behavior is essential for optimizing performance and safety. However, building a full-scale experimental model is often impractical due to high costs and the reactor’s large size. A scaled-down model provides an efficient and cost-effective approach to studying critical aspects of fluid flow, heat transfer, and pressure distribution while capturing key physical phenomena. This study presents the design and computational fluid dynamics (CFD) validation of a 1/5 scaled-down mock-up of the MSRE. The scaled-down model was developed to replicate the geometry of the original MSRE while maintaining fluid behavior, such as Reynolds number. The study also provides a detailed explanation of the scaling laws used to ensure that the down-scaled model accurately reflects the behavior of the full-scale system.

Given the reduced size, the model cannot replicate every detail, such as all the surfaces and channels exposed to molten salt within the reactor core, making validation crucial. CFD simulations were performed using the scaled model to analyze fluid flow and pressure characteristics. The results of the simulations were compared to experimental data from the original full-scale MSRE. This comparison confirmed the accuracy of the scaled mock-up and its reliability for predicting the thermohydraulic behavior of molten salt reactors, making it a valuable tool for further research.

Differentially-heated cavity natural convection is an important phenomenon relevant to the design of thermal energy storage systems, concentrated solar power receivers, building-integrated photovoltaic systems, and nuclear reactor passive safety systems. Particle Image Velocimetry (PIV) measurements are implemented to study the natural convection behavior of molten nitrate salt in a differentially-heated cavity for Rayleigh numbers up to 10⁹ and Prandtl numbers from 22 to 30. A low melting point salt mixture, NaNO₃-KNO₃-LiNO₃-CaNO₃, is selected as a working fluid to provide operating temperatures from 100°C to 500°C. The experimental methodology for PIV measurements in a heated molten salt test section with a transparent optical window is presented along with preliminary test data. 2-D planar PIV measurements of the flow field in molten nitrate salt are compared to measurements of the flow field in water, with matching Rayleigh numbers and Prandtl numbers from 2 to 7. Quantification of measurement uncertainties is described and compared to alternative flow measurement techniques.

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

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

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

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

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

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

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

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

In the case of a LOCA in the PWR, during the long-term cooling stage, boric acid in the coolant may be carried with steam or entrained droplets due to steam discharge, which may affect the reactivity of the core. There are a number of research that make it possible to determine the value of droplet entrainment leaving the reactor. However, the loss of boric acid with steam is related to the solubility at steam. It is necessary to study dissolution process of boric acid in steam under different parameters.

This paper conducts an experimental study of the distribution coefficient of boric acid between the steam and liquid phases of the solvent. The test device of the solubility of boric acid with steam has been established, including a heated test section, steam separator, and steam condenser. The experiments are conducted under a certain pressure of 0.1-0.4MPa, with an initial boric acid solution concentration range of 1000-10000 ppm. The concentration of boric acid in the discharged steam and the remained solution are obtained. The effects of pressure, temperature and initial concentration on the solubility of boric acid in steam are summarized. The empirical relationship of distribution coefficient of boric acid in the steam phase and liquid phase are obtained by fitting the experimental data.

This study investigates the flow characteristics in a 3x3 Helical Cruciform Fuel (HCF) bundle using Magnetic Resonance Velocimetry (MRV) experiments and Computational Fluid Dynamics (CFD) analysis. The HCF bundle, designed to enhance thermal-hydraulic performance in nuclear reactors, features a unique geometric configuration with helically twisted cruciform-shaped fuel elements. To validate the numerical predictions, experimental measurements were conducted using MRV technology, which provides three-dimensional velocity field data without intrusive flow disturbance.

The experimental facility consisted of a full-scale 3x3 HCF bundle model operating at high Reynolds numbers exceeding 10,000. MRV measurements captured the complex flow structures, including secondary flows and vortex formations in the sub-channels. The CFD analysis employed the Reynolds stress model with a refined mesh containing approximately 2.5 million elements, validated through rigorous mesh sensitivity studies.

The study revealed distinct flow patterns characterized by enhanced mixing due to the helical geometry. Secondary flows were particularly pronounced in corner sub-channels, exhibiting higher tangential velocities compared to interior sub-channels. These findings provide crucial validation data for CFD methodologies in nuclear fuel bundle analysis and contribute to understanding the thermal-hydraulic behavior of advanced fuel designs.

The validated CFD model can serve as a reliable tool for future HCF bundle optimization studies and thermal-hydraulic characteristic analyses, potentially leading to improved nuclear reactor fuel efficiency and performance.

Thermal fatigue cracking may initiate at a T-junction where high and low temperature fluids flow in. In this study, the flow structure and fluid temperature fluctuations in a branch pipe of a T-junction were investigated under flow patterns where the main pipe flow penetrates into the branch pipe. The test section consists of a horizontal main pipe with an inner diameter of 150 mm and a vertical branch pipe with an inner diameter of 50 mm. A 45º elbow was installed upstream on the branch pipe side in order to study its effect on the penetration flow pattern and temperature fluctuations. To simulate penetration flow, the experiment was conducted under conditions where the inlet flow velocity on the branch pipe side was much smaller than the inlet flow velocity on the main pipe side. The flow pattern was visualized using the tracer method. The flow in the branch pipe was classified into three flow patterns: no penetration; entrained penetration; and impinged penetration. These patterns depended on the momentum ratio of the main and branch pipes, regardless of the presence of the elbow. The maximum penetration depth into the branch pipe increased when the upstream elbow was installed. Fluid temperature distribution along the branch pipe was measured with eight sheathed thermocouples. The fluid temperature fluctuations also increased, especially in the range of relatively small momentum ratios where the hot mainstream intermittently penetrated into the branch pipe.

Fatigue and defect tolerance assessments of high integrity PWR pressure boundary components require transient temperature fields to be defined to produce thermal stress predictions. These temperatures are often produced using a heat transfer coefficient and bulk temperature boundary condition approach. This can be imprecise and inaccurate for components with complex flows and geometries. An alternative approach is to directly predict the temperatures using conjugate heat transfer CFD, where the solid temperature field is predicted directly and simultaneously with the adjacent flow. This approach removes the uncertainties of using an intermediate model to transfer the information, but since it requires flow predictions at all times, the computational cost is impractical for the large number of multi-hour plant transients that need to be considered.

The cost of the CFD solution can be reduced by using an 'infrequent updates' approach, where the flow-field is considered to change slowly and is 'frozen' for intervals where only the thermal fields are solved. This is cheaper to calculate and can use longer time steps. The flow is solved in brief update intervals throughout. This method has been applied to the assessment of a large number of transients for the feedwater nozzles for the steam generators at the Sizewell B PWR in the UK.

The setup considerations and accuracy of the infrequent updates approach are discussed, as well as the effects of finite domain size and buoyancy driven flow instabilities.

Hydroaccumulators are designed to provide fast water injection into the vessel to ensure proper core cooling. One of its main function is that during LOCA cases, when no HPSI pumps are available, the hydroaccumulator supplies cooling water until LPSI pumps can operate and maintain the long term cooling. In VVER-440 reactors the hydroaccumulator empties at higher pressure than the LPSI can operate. The core should survive that injection hiatus. In that sense the primary pressure when hydroaccumulator injection terminates is very important

In PMK-2 facility several experiments were conducted utilizing various ECCS configurations and these tests were later calculated using RELAP5 thermal-hydraulic system code. During these post-test calculations a difference between the measured and calculated hydroaccumulator injection characteristic and terminating pressure value was noticed.

The hydroaccumulator is pressurized using nitrogen gas. Under pressure, some of this gas gets dissolved into the coolant in the hydroaccumulator water. During injection, the dissolved gas is reintroduced into the gas dome increasing its pressure. The RELAP5 system code does not consider this effect, leading to the observed differences. This is an unconservative deviation since the code predicts lower pressures at HA emptying thus shorter injection hiatus.

To address the phenomena experiments were performed. Using the PMK-2 facility more than 70 separate effect tests were conducted using one of the hydroaccumulator vessels. The tests were done at several different pressure levels and coolant temperatures. RELAP5 post-test calculations were carried out for each test, and the effect of introducing additional gas into the vessel was studied.

A significant percentage of reactors in operation, under construction or recently commissioned are VVER reactors. In parallel, there is a growing interest in analyzing the behavior of the Advanced Technology Fuels (ATF) under development in this type of nuclear reactor. Among the new ATF designs, the FeCrAl and Cr-coated claddings are the most promising evolutionary options. In addition to a relatively high level of technology readiness, these evolutionary cladding materials offer improved oxidation and hydride resistance, as well as improved mechanical strength. All these properties are essential in accident sequences where high core temperatures are reached. In the present study, core damage sequences have been analyzed with a model of a VVER-1000 reactor for the thermal-hydraulic code TRACE. In addition, an in-house version of the TRACE5P6 system code for FeCrAl cladding has been developed by NFQ and UPM.The selected sequences are SBO and LOCA sequences. The results show that the core damage temperature for the Zircaloy cladding cases is reached well before that for the FeCrAl and Cr coating cladding cases. The performance of these new cladding materials provides additional time for recovery of the safety systems responsible for core cooling and replenishment of the reactor coolant system inventory.

The motivation of the study of fully-implicit scheme for two-fluid model is, comparing with semi-implicit scheme, the mainly advantage of fully-implicit method is the restriction of the time step on stability is very small, and large time step is allowed. However, fully-implicit scheme makes the degree of coupling among equations is strong, and the solution become more difficult. First, the calculation of Jacobian matrix is difficult and highly time- and memory- consuming due to matrix is large. The Broyden-type method is adopted because it only requires calculate Jacobian matrix one time during the iteration process. We use numerical difference to estimate Jacobian matrix for the first iteration, then calculate Jacobian matrix by Broyden method for the remaining iterations. Due to the sparsity of Jacobian matrix, the Schubert method is applied. This method makes full use of Jacobian matrix’s sparsity. Second, the convergence performance become poor especially for strong interfacial effect case. The reason is the non-linear interfacial models makes the Jacobian matrix highly ill-conditioned. The relationship between condition number and the degree of interfacial effect is studied, and we try to reduce condition number by modify governing equation. Finally, this solver is tested. The calculation performance under very larger time step is assessed.

High-resolution simulations of two-phase flows can augment existing experimental data used for thermal-hydraulic system code development. PHASTA is one such high resolution code that uses Direct Numerical Simulation of the Navier-Stokes equations with the Level Set method to resolve individual bubbles and droplets of the flow. PHASTA, when deployed on high performance computing systems, has shown remarkable performance at simulating turbulent bubbly flows which occur in prototypical reactor geometries and conditions. However, flows of high void fraction systems present a challenge to the Level Set method which is well known for its mass loss deficiencies. High void fraction annular flows are typical near the top of boiling water reactor fuel channels and steam separators. Annular flows generate a large amount of small, entrained droplets and bubbles which require a prohibitively fine mesh in order to resolve and conserve the mass of these smaller objects. New numerical methods or models must therefore be incorporated into PHASTA in order to accurately model these flows with economical grid sizes. The Conservative Level Set method is one such method which has shown superior mass conservation properties on a variety of complex two-phase flows. This work describes the development and testing of the Conservative Level Set method implemented in the PHASTA finite-element code. A simulation of a swirling annular flow in a BWR steam separator is conducted to test the method on a large engineering-scale problem.

An important industrial issue and a continuing challenge in computational fluid dynamics (CFD) is the simulation of gas-liquid flows when several two-phase regimes coexist.

To overcome the computational challenges associated with all-scale interface resolving approaches, all-flow-regime CFD models have been proposed. One such method is the Generalized Large Interface Model (GLIM), implemented in the NEPTUNE_CFD software. This framework allows a smooth modelling approach transition: the small-dispersed scales are modelled, whereas the large scale gas-liquid interfaces are explicitly treated. These methods can be useful for many nuclear thermal-hydraulic applications.

To achieve this goal, GLIM and other models in the literature require specific closure terms for configurations where the gas-liquid interface is recognized. In addition to the interfacial closure formulations, blending functions and interface recognition methods are essential to deal with the transition between scales.

The aim of the present work is to evaluate different formulations of the interfacial forces. The NEPTUNE_CFD code is used. A rising large bubble configuration where the interface can be well resolved or rather diffused over several mesh cells was considered as a preliminary validation test. Several large interface drag formulations and cell number over bubble diameter ratios have been tested, the results are discussed in this paper. Then, the model was applied to simulate an intermittent air-water cross-flow in a tube array, featuring at the same time dispersed bubbles/droplets and large gas-liquid interfaces, a configuration close to the one encountered in U-tube steam generators.

Pool scrubbing is an effective process to decontaminate radioactive aerosols as severe accidents happen in nuclear power plants. Bubble residence time is one of the key parameters to determine the aerosol decontamination factor (DF) which is defined to describe the efficiency of aerosol removal, especially in the swarm flow region which makes a significant contribution to the total aerosol removal. The Euler-Euler two-fluid method and the Lagrangian Particle Tracking (LPT) method are applied, the former is used to get the flow field information, and the LPT method is used to track the bubble movement to obtain the bubble residence time. Through the analysis of bubble residence time distribution, the model of probability density function for bubble residence time is preliminarily established. The probability density function obeys well an exponential decay behavior. The decay constant is fitted according to the CFD simulation results. In general, the developed model shows good potential in predicting bubble residence time. Furthermore, the effect of bubble diameter on the probability density function is investigated. The bubble diameter shows a strong effect on bubble residence time, which is because larger diameter bubbles experience greater buoyancy, resulting in a higher rising velocity.

An ex-vessel core catcher for advanced light water reactors is being developed to stabilize molten corium outside the reactor vessel and prevent molten corium concrete interaction (MCCI) during severe accidents. The core catcher includes a sacrificial concrete (SC) layer, carbon steel body, protective material, and an external cooling channel. Its key function is to collect molten corium and remove heat, while preventing re-criticality and excessive hydrogen generation. During severe accidents, molten corium discharge and some parts of a reactor pressure vessel (RPV) are expected to impact directly on the core catcher body. Given their high momentum and large internal energy, it might degrade the integrity of the structural bodies, affecting safety performance of the core catcher. To investigate these scenarios, we developed a smoothed particle hydrodynamics (SPH) framework to analyze the interaction among the molten corium, reactor pressure vessel, and core catcher. As a fully Lagrangian particle-based method, SPH is suitable for handling free surface of corium flows and structural topology changes in the RPV and core catcher body. The explicit incompressible SPH (EISPH) model simulates corium behavior with an implicit viscosity solver for computational efficiency. A finite multiplicative model based on Total Lagrangian SPH (TLSPH) handles large-strain elastoplasticity in the structures. The coupled EISPH-TLSPH model is used to investigate mainly corium relocation behaviors and structural integrity under several potential accident scenarios. We expect that the proposed model will be useful for accurate safety analyses of the accident mitigation strategy using the ex-vessel core catcher.

This paper presents the analysis of the effect of Molten Corium Stratification (MCS) during Molten Corium Concrete Interaction (MCCI) using developed the PUMBAA (Prediction modUle of MCCIs with Basemat Attack and Ablation) code under the severe accident analysis platform called SAFARI (Safety Analysis Code For Severe Accident Risk Identification) currently being developed in Korea. PUMBAA built upon CORQUECH developed by Argonne National Laboratory in US, as a reference code focuses on the MCCI phenomena, aiming to extend its ability to analyze MCCIs phenomena interlaced with Molten Corium-Water Interactions (MCWIs) and Containment Thermal Hydraulics (CTHs) along with Severe Accident Management (SAM) actions during severe accidents. In this study, the development and performance of PUMBAA’s corium stratification analysis capability are introduced and validated against the MOCKA 3.1 experiment. Part of the MOCKA series, this experiment simulates a dry 2D cylindrical siliceous concrete cavity, where continuous corium pouring induces various physical phenomena. The paper first describes the modeling of corium stratification in PUMBAA and key physical phenomena in the MOCKA 3.1 experiment. It then presents the validation process and result analysis. The results show that the PUMBAA code, through its corium stratification analysis, effectively accounted for the physical phenomena of the stratified oxide and metal layers observed in the MOCKA 3.1 experiment, accurately predicting the trends in the experimental data.

The melt jet breakup is an important phenomenon for assessment of the ex-vessel phase severe accident, which is highly related to the debris bed coolability. The violent two-phase boiling is accompanied by the melt jet breakup phenomenon due to the high temperature of the melt, and it can affect the jet breakup behavior. The effect of the vapor generation intensity, which represents the two-phase mixing zone behavior, was investigated by controlling both the melt and water temperatures. The melt jet breakup length was observed by visualization using high speed cameras. Based on the experimental observations, the effect of the vapor generation intensity was confirmed. As the vapor generation intensity increases, the jet breakup length became longer. Therefore, the parameter for the vapor generation intensity was suggested to develop the new jet breakup length correlation including the vapor generation intensity parameter so that the existing correlations (Saito correlation and Epstein & Fauske correlation) could be integrated into single correlation.

Hualong One is the first independently developed million-kilowatt-class pressurized water reactor nuclear power plant in China, which meets the design standards of the third generation nuclear power technology. The Cavity Injection and Cooling System (CIS) for the reactor vessel is one of the measures to mitigate severe accidents in Hualong One, and its unique combination of active and passive technologies can effectively prevent the rupture of the reactor pressure vessel and achieve the retention of molten debris within the reactor. Based on the layout features of the CIS system in Hualong One and the design experience of the first unit, this paper proposes design optimization solutions and improvement measures from the perspective of layout design for subsequent projects, improving the compact arrangement of pumps and valves in the plant, and providing valuable design experience for future PWR nuclear power plant design in China.

In a postulated station blackout accident, the fuel channels inside of a Canada Deuterium Uranium (CANDU) reactor would dry out, heat up, and collapse to the calandria vessel bottom. Due to decay heat generation, the debris bed would continue to heat, compact, and melt, forming a molten corium pool. Here, the transient heating and melting of a compacted debris bed is simulated using unsteady Reyolds-averaged Navier-Stokes based computational fluid dynamics.

A time-varying decay heat is used with the starting conditions representing post moderator dry-out. A source-based enthalpy-porosity phase change model is employed to capture the non-eutectic melting process, accounting for a 500K difference between the solidus and liquidus temperatures of the corium. The developing molten region, characterized by a maximum modified Rayleigh number around 10¹², is modelled with the k-ω turbulence model. Turbulence is allowed to develop from an imposed very low level in the melting region consisting of a growing liquid pocket and a partially molten layer, as the temperature locally exceeds the solidus temperature.

Heat flow through the vessel wall to the surrounding water is modelled with a conjugate boundary, and a convection-radiation boundary is applied to the corium top surface. Verification and validation cases are done based on previous studies using molten-salt corium simulants.

The evolution of the molten corium velocity and temperature fields, the unmolten crust thickness, as well as their impact on the exiting heat flux are presented and analyzed. These findings assist the in-vessel retention studies of CANDU reactors and inform future modelling efforts.

Hydrogen distribution in the reactor building of Fukushima Daiichi NPP Unit 3 during the accident was analyzed using the CFD code BAROC. The main feature of BAROC is that it solves the pressure Poisson equation based on energy conservation. This makes the calculation stable and fast, even under sudden changes of fluid conditions. The total number of spatial meshes with 50 cm cubic was approximately 750,000. Transient from the hydrogen inflow to 18 hours later were analyzed in 14 parallel using an Intel Gold 5218 2.3GHz CPU, and the computational time was almost real time. Initial conditions were assumed to be that the building was filled with air at room temperature and atmospheric pressure. Hydrogen was assumed to have entered the building from a shield plug on the 5^th top floor of the building. The analysis assumed that 75 tons of steam and 650 kg of hydrogen would both enter the building. Although the blowout panel in the 5^th floor was not opened at the accident, it was assumed that there was a certain amount of leakage because the building was not a leak-tight structure. Analyses were performed for 3 cases with 2%, 3.3%, and 6.6% leakage of the blowout area. The analysis showed that the hydrogen concentration in the 5^th floor was within the flammable range regardless of the leak area. It was also found that when the amount of inflow hydrogen was increased to 1,300 kg, the hydrogen concentration was within the detonation range in the 5^th floor.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

The use of molten salt reactors (MSRs) presents a promising avenue for achieving energy independence and reducing reliance on fossil fuels. A key challenge in MSR development is enhancing heat exchanger efficiency while minimizing pressure drop and operational costs.
This study experimentally investigates the thermal-hydraulic performance of additively manufactured helically rifled tubing, a passive heat transfer enhancement (HTE) method that promotes flow mixing through centrifugal forces. The pressure drop characteristics of helically rifled tubes with different twist pitches (y = 3 and y= 6) were measured and compared to plain tubing, alongside computational fluid dynamics (CFD) simulations. Using Buckingham-Pi analysis, nine dimensionless groups were identified to characterize flow behavior and thermal performance. Preliminary experimental results indicate that existing helical rifling correlations significantly under predict frictional losses by 10%, highlighting the need for refined modeling approaches. A strong agreement was observed between simulations performed by Vulcanoff et al. and the experimental friction factor data for the helical rifling at the same geometric parameters (y=3). Geometric variation of the twist pitch from y=3 to y=6 suggests the dependence of the nondimensional parameter y/d_min diminishes at sufficiently high Reynolds numbers. Computational findings by Vulcanoff et al. suggest that helically rifled tubes improve heat transfer efficiency. Investigations are ongoing into the Nusselt number and thermal performance factor. This work provides a foundation for optimizing helical rifling geometries in MSRs, contributing to advancing advanced nuclear energy systems.

A new phenomenon, namely tangential oscillation of thermal stratification, can cause periodic temperature changes on the pipe wall and leads to material damage of thermal fatigue. At North China Electrical Power University, an experimental facility has been constructed and operated to investigate the initiation mechanism of the tangential oscillation of thermal stratification at pipeline elbows. In this study, experiments have been performed with variations of elbow-radiuses and inlet flow temperature. Results show temperature increase at the intrados side of the elbows, which indicates an angular shift of the thermal stratification at the elbow due to the centrifugal force. Thermocouples downstream of the elbow have captured temperature changes in the near-wall flow. The elbow-radius shows a clear influence on the locations of the high temperature region in the thermal stratification. In addition, temperature fluctuations have been calculated based on the measurement data. The location with the maximum temperature fluctuation can be found in the place, where the mean temperature reaches the maximum. Moreover, frequency spectra of the temperature data do not show any significant peak. Combined with the calculation results of Richardson-number, it can be understood that the thermal stratification is not stabile enough to keep the tangential oscillation downstream of the elbow. It leads back to the turbulent mixing enhancement at the elbow, which is clearly increased with decrease of elbow-radius. However, decrease of the elbow-radius leads to increase of the temperature fluctuation in the near-wall flow, which indicates a higher potential of thermal fatigue.

The Small Modular Reactors (SMRs), as an emerging nuclear energy technology, hold great promise for gradually replacing coal-fired power plants in the future due to its flexibility, high safety, and economic advantages, thereby contributing to the decarbonization of energy systems. In a domestically developed integrated SMR in China, the steam generator adopts a helical heat transfer tube design, with the outlet steam being saturated steam. Considering the compact size and high-level power density characteristics of SMRs, it is necessary to design an efficient and compact moisture separation component, to ensure that the quality of steam entering the turbine meets the required standards. This study conducted cold-state experiments and theoretical analyses on a designed swirl vane gas-liquid separator. Under conditions with different drainage step heights, the critical separation boundaries for both low-wetness and high-wetness scenarios were determined. For steady-state conditions, based on experimental data of the separation efficiency and pressure drop of guide vanes with different profile variation patterns, a predictive correlation was proposed. For unsteady conditions, a predictive model was developed to describe the transition from swirl annular flow to churn flow, and a flow regime map was constructed by integrating extensive experimental data with previous studies. These findings provide important theoretical support and experimental evidence for the optimization and performance enhancement of SMR steam-water separation systems.

Globally, advanced reactors are adopting helical tube steam generators to reduce volume compared to traditional U-tube designs in large light water reactors. These steam generators feature a once-through operation and a shared header for multiple heat transfer tubes but often encounter issues associated with two-phase flow instabilities such as Density Wave Oscillations (DWO). Such instabilities are critical in boiling water reactor cores and pressurized water reactor steam generators, causing significant flow and pressure oscillations, which can potentially degrade the integrity of a system.

This study aims to address this gap by comparing experimental results on two-phase flow instabilities in helical tubes from previous research works with the predictions obtained from Korean nuclear safety codes MARS-KS and SPACE. The objective is to assess whether the current helical tube thermo-hydraulic models in these codes adequately reflect observed physical behaviors or if there are significant discrepancies that require model enhancements. This analysis intends to provide insights into the dynamics of two-phase flow in helical steam generators and help improve the predictive accuracy of safety codes, thereby enhancing the reliability and safety of advanced nuclear reactor.

In nuclear power systems, the insertion of a throttling device at the inlet of the heat exchange tubes in steam generators enhances the flow resistance within the single-phase flow region of the tubes, thereby mitigating the risk of flow instability within the steam generator. This study proposed a novel gear-type throttling device designed specifically for steam generators. Various gear-type throttling prototypes with differing gear heights were designed and fabricated for experiment and numerical analysis .Through a systematic experimental testing and numerical simulations, the resistance characteristics of the throttling device with different structural parameters were obtained in a wide range of flowing conditions, . The results reveal that the resistance coefficient of the innovative gear-type throttling device can fit for different operational requirements in steam generators. The resistance coefficient exhibits significant sensitivity in gear height and width. Additionally, the resistance coefficient for throttling devices with varying gear heights remains relatively stable across different Reynolds numbers.A mathematical relationship was established to correlate multiple structural parameters and the resistance coefficient. This work is valuable for the design optimization and validation of next-generation steam generators in the nuclear energy system.

The function of reactor core residual heat removal system is to extract heat from the core and reactor coolant systems during the shutdown of the power plant. According to the principle of simplified configuration, improving safety and economy, possible reactor core residual heat removal systems are studied, and the configuration of recommended optimization solution is fixed.The optimized configuration is based on two completely independent series; This configuration can be used for core residual heat removal function after connecting the primary circuit, and for containment spray function. the analysis of single failure and system capacity for this optimized configuration are performed. The analysis of single meets the safety requirements, except that the exemption criterion is used for the mode of residual heat removal, which can make sure the nuclear power plant is brought to the safe state.The capacity of key equipment including pumps and heat exchanger is analyzed,which is in making full use of the equipment of original some reactor, considering the equipment design, the shutdown time of plant, the design limit of containment and layout space comprehensively, the actual shutdown time and the pressure and temperature of containment are calculated, finally the capacity of equipment can meet the mode of residual heat removal and the mode of containment spray. The optimized reactor core residual heat removal scheme, not only improves the safety of the power plant, but also improves the economics, which has the great significance to the subsequent improvement of the market competitiveness of power plants.

Due in great part to the earthquake and ensuing tsunami, the Great East Japan Earthquake of 2011 seriously damaged the Fukushima Daiichi Nuclear Power Plant and resulted in a major accident. The malfunction of the cooling water level measuring system was one element causing this accident. Differential pressure gauge monitoring of the reactor pressure vessel (RPV) water level was used at the time. But the temperature of the reference side piping surged greatly during the severe accident, which caused the water on the reference surface to evaporate. It is not possible to precisely identify the real drop in water level since this evaporation lowered the differential pressure between the water level inside the reactor containment vessel and the reference piping.

In order to find a solution to this issue, we are working on a new water level meter that is capable of functioning even in severe accidents. Through the utilization of an ultrasonic transducer (TDX), this apparatus enables real-time measurements to be taken from outside the containment vessel of the reactor. The concept of measurement originates from the disparity in the reflection coefficients of ultrasonic waves that travel through metal that is in contact with water as opposed to air.

The results of experiments measuring the reflection coefficients of metal walls in contact with air and water using a small water level measuring device are reported in this work. Comparisons with hypothetical values computed with the ultrasonic wave propagation simulator SWAN21 verified the validity of the experimental results.