Thermal mixing and stratification in large pools and enclosures play a critical role in the safety and performance of pool-type nuclear reactors, particularly during transient scenarios involving significant temperature differences between incoming and bulk coolant. Accurate modeling of these phenomena is essential for predicting system behavior and supporting passive safety features such as natural circulation. This paper presents a new 1D model for thermal mixing and stratification, developed and implemented in the SAM code. The model represents a large pool as 1D coolant jet channels and zero-dimensional bulk pool volumes, enabling the simulation of a wide range of flow configurations, including hot and cold jet interactions, stratified layers, and the influence of complex geometries such as ceilings, free surfaces, and internal obstacles. Heat exchange between jet and pool regions is governed by closure relations calibrated against 3D computational fluid dynamics (CFD) simulations. The model improves upon earlier approaches by incorporating time-dependent jet characteristics and capturing the associated delay effects more accurately. Code-to-code comparisons and validation against experimental data from the Thermal Stratification Test Facility demonstrate the model’s accuracy and flexibility. This work offers two key contributions: (1) an efficient and robust method for simulating thermal mixing and stratification at the system level, eliminating the need for external coupling between system analysis codes and CFD, and (2) a significant enhancement of SAM’s capabilities to analyze thermal stratification phenomena in advanced reactor systems.

— The U.S. Department of Energy’s Microreactor Program, along with Idaho National Laboratory’s development of a nuclear microreactor applications platform named MARVEL, support research and development for deploying small, transportable reactors across civilian, industrial, and defense sectors. The MARVEL microreactor, an 85-kW(thermal) fission reactor, incorporates TRIGA nuclear fuel and a sodium-potassium eutectic as its primary coolant. It is designed for safe and efficient operation using natural circulation and a low power density core. The reliance on natural circulation for primary cooling means the reactor avoids using fuel spacers to minimize core pressure drop, which could disrupt the primary coolant’s natural flow. However, the core’s tight pitch-to-rod diameter ratio of 1.056 raises the risk of fuel rod contact and elevated peak cladding temperatures. To ensure reactor safety, this study conducted computational modeling and simulations to examine the reactor’s thermal-hydraulic-mechanical characteristics, including the reactor core heat transfer coefficients, potential for rod-to-rod contact, and its impact on peak cladding temperature. The analyses indicated that the thermal deformation of the fuel rods under a worst-case scenario may lead to fuel rod contact, but peak cladding temperatures remained well below safety criteria, ensuring safe reactor operation without fuel spacers under steady-state normal operating conditions.

The migration of neutrons from their generation point to the absorption point in a light water was analyzed using the Monte Carlo method. Neutron age (NA) of fast neutrons and the diffusion area (DA) of thermal neutrons were calculated as functions of temperature (20 – 100 °C), pressure (2 - 10 MPa), and boron concentration (200 - 1000 ppm), as well as their combined effects. NA and DA at room temperature and pressure have been calculated, and the results are consistent with experimental values. The percentage changes in average NA and DA per unit increase of boron concentration, temperature, and pressure have been estimated. In all cases, the amount of DA change was larger than that of the NA change. The percentage decreases of average NA and DA per fixed amount increment of temperature, pressure, and their combined increment were evaluated with 1000 ppm boron in the medium. Maximum average NA and DA decrements were 2.7% and 64.4%, respectively, at room temperature. This investigation provides critical insights into reactor safety and efficiency by quantifying how boron concentration, temperature, and pressure affect neutron migration in light water systems using MCNP. Bangladesh Journal of Physics, Vol. 32, Issue 2, pp. 53 – 65, December 2025

The Microreactor Applications Research Validation and Evaluation (MARVEL) microreactor utilizes natural circulation as a core cooling mechanism and liquid metal as a primary coolant. Moreover, the reactor core has a pitch-to-diameter ratio of 1.054, which is considered a tight lattice configuration. Numerous studies have widely reported that Reynolds-averaged Navier-Stokes (RANS) turbulence models inaccurately predict heat transfer in liquid metals and fail to capture flow pulsations that can occur within tight lattices, leading to further inaccuracies in simulation results. Therefore, evaluating the accuracy of RANS turbulence models in the thermal-hydraulic analysis of the MARVEL microreactor core is crucial for assessing reactor safety. In this study, a large eddy simulation (LES) of the MARVEL microreactor core subchannel was conducted and compared with a RANS simulation to evaluate the accuracies and conservatism of the RANS model. In comparison with the RANS model, LES captures flow pulsations in a tight lattice that enhance heat transfer, whereas the RANS model underpredicts heat transfer in liquid metal flow. As a result, the RANS model predicts a higher peak cladding temperature than LES. However, owing to the high thermal conductivity of the liquid metal, the discrepancy between the two approaches is limited. These results indicate that the steady-state RANS model is adequate for the thermal analysis of the liquid metal–cooled MARVEL microreactor core and can provide conservative, yet not excessively overpredicted, results for safety assessment.

Foreword: Special Issue Featuring Papers from the 2024 Advanced Reactor Safety Topical Meeting (ARS 2024)

Conventional inductance control rod position measurement sensors (ICRPMS) still suffer from insufficient measurement accuracy and the inability to achieve continuous measurement, which seriously affects reactor safety. Thus, a novel ICRPMS based on a segmented coil structure is proposed in this paper. To investigate the factors influencing the output characteristics of the sensor, an analytical model with a driving rod based on the truncated region eigenfunction expansion (TREE) method is established, and the feasibility of the model is verified by simulation and experiments, showing relative errors of less than 6.2% and 6.7%, respectively. Subsequently, to enhance integration of the measurement system, a single-phase inverter circuit is designed as the excitation source, ensuring a stable sinusoidal output voltage. Finally, the precision is evaluated under cold and thermal states. Experimental results indicate that the maximum error of the sensor in continuous measurement mode is within 2 mm, and the maximum nonlinearity error is less than 1.2% over the full 320 mm stroke without compensation.

Nuclear power plants can provide CO2-free electricity and heat, supporting the European net-zero emissions objectives. The Polish energy transformation strategy includes the development of large-scale pressurized water reactors, as well as small modular reactors (SMRs). Both state-owned and private companies consider SMRs as a potential source of electricity and beyond, i.e. district heating, industrial process heat, and hydrogen production. Over the past few years National Centre for Nuclear Research, Poland (NCBJ) has been involved in several high temperature gas-cooled reactor (HTGR) projects at both the national (Gospostrateg-HTR and HTR-MEiN) and European level (Gemini Plus and Gemini for Zero Emission). At the national level, a small-scale, prismatic-type research HTGR of 30 MW(thermal), named the HTGR-POLA, is being considered to be built at the NCBJ site. Its main mission would be to serve as a demonstrator of HTGR technology for Polish industry. This paper investigates the reference HTGR-POLA plant design (v1), the resulting optimized core configuration (v2), and their impact on the reactor safety performance during selected design-basis accidents (DBAs). The proposed optimization of the core, from the initial to the optimized configuration, was successful, with the aim of improving the overall safety of the plant. The introduced capability of the coupled neutronic/thermal-hydraulic phenomena simulations for the reactor core over the whole fuel cycle is a promising approach, as it allows for the approximation and identification of the most relevant safety issues without relying on overly conservative assumptions. The results of the calculations performed using the MELCOR 2.2. code are presented for selected accident scenarios, depressurized loss-of-forced circulation and pressurized loss-of-forced circulation, which represent different types of postulated initiating events for HTGRs. Specifically, these scenarios correspond to a pipe break and a primary blower stop, respectively. The presented core volume fraction, as a function of fuel temperatures, shows that during most severe scenarios, the fuel temperature remains more than 500°C below the safety limit of 1600°C, at which the fission product release rate from the fuel significantly increases. These results, which will support the future probabilistic safety analyses of the HTGR-POLA reactor by assessing the accident consequences, will constitute a significant component of the preliminary safety analysis report (PSAR). The PSAR is a regulatory and legal requirement for newly built nuclear power plants and research reactors in Poland.

Organisations in the Czech Republic are involved in international research and development of advanced helium-cooled nuclear reactors, including both the very-high-temperature reactors (VHTRs) and the gas-cooled fast reactors (GFRs). To support this effort, a dedicated research infrastructure has been developed and constructed, incorporating large-scale facilities, such as the High-Temperature Helium Loop (HTHL), the S-Allegro helium loop, and other specialised equipment. Current studies are investigating the resistance of structural materials in high-temperature helium environments. Various metallic alloys and ceramic materials intended for high-temperature applications are tested at 750–900 °C. Additional activities focus on helium-coolant technologies – particularly purification, purity monitoring, recovery, and primary-circuit sealing – and on reactor safety and system behaviour under off-normal and emergency conditions.

Accurately characterizing the temperature dependence of UO2 thermal conductivity is crucial for evaluating its performance under nuclear reactor operating conditions. However, experimental measurements are costly, density functional theory (DFT) calculations are constrained by small spatiotemporal scales, and traditional empirical potentials struggle to capture strong anharmonic effects. To this end, we developed a machine-learned neuroevolution potential (NEP) with near-DFT accuracy using an active learning strategy, and we systematically evaluated and cross-validated the thermal conductivity of UO2 using equilibrium molecular dynamics (EMD), homogeneous nonequilibrium molecular dynamics (HNEMD), and nonequilibrium molecular dynamics (NEMD). The results demonstrate that HNEMD delivers a high signal-to-noise ratio, low uncertainty, and rapid convergence, exhibiting superior computational efficiency and robustness. At 500 K, the spectral phonon mean free path spans approximately one order of magnitude, and heat-transport channel lengths exceeding about 5 µm approach the bulk thermal conductivity limit. In the 800-1500 K range, the NEP reproduces the experimental temperature dependence of UO2 thermal conductivity, while at lower temperatures (300-800 K), it achieves predictive accuracy comparable to that of DFT+U. Systematic validation of UO2 fundamental properties including the equation of state, phonon dispersion relations, elastic constants, heat capacity, and linear thermal expansion coefficient demonstrates that the constructed NEP is reliable and broadly applicable. This work provides methodological support for multiscale thermal transport modeling of nuclear fuels and reactor safety assessment.

Accurate transient behavior modeling is crucial for assessing the safety of sodium-cooled fast reactors (SFRs) . This study validates the ISAA-Na code through simulations of the CABRI-BI1, AGS0, and E7 experiments. The CABRI-BI1 experiment, focused on sodium boiling, was simulated using a multi-bubble slug flow model. Results showed that ISAA-Na more accurately predicted coolant temperature and pressure prior to boiling than SAS4A, ASTEC-Na, and SIMMER. In the CABRI-E7 experiment, ISAA-Na’s cladding failure model was tested under transient overpower (TOP) conditions. ISAA-Na’s results closely matched experimental data and other codes, successfully predicting fission gas release, coolant temperature, and cladding failure timing and location. The CABRI-AGS0 experiment further examined fuel pin behavior under TOP conditions. ISAA-Na accurately captured fission gas release, temperature, and fuel melting radius predictions. Overall, ISAA-Na demonstrates strong potential for SFR transient analysis, though further refinement in phase transition modeling is needed. This validation provides valuable insights into ISAA-Na’s capability to enhance SFR safety assessments.

An integrated open-source CFD workflow for gas distribution and flame propagation analysis in nuclear reactor safety

Abstract Understanding two-phase flow is key to the safety and efficiency analyses of light water nuclear reactors (LWRs). The bubble departure diameter highly influences the prediction of the heat flux of the system. By using high-speed photography, this work seeks to conduct a quantitative examination of individual forces contained in a force-balance approach for bubble departure diameter prediction. A comparison of different models to estimate the growth force can provide further insight into the force-balance approach. Computation of an experimental growth force had an average magnitude of ≈10−6 N. However, when using the Rayleigh–Plesset and added mass models with the radius functions used by Klausner et al. (1993) and Sugrue and Buongiorno (2016), the average magnitude for the growth force was higher at ≈10−3 N, while at the same time the profile shape of these modeled growth forces differed from that of the experimental growth force. In comparison, when using the growth force models with the experimental radius, the profiles of these forces were similar to the experimental growth force profile shape, but the average magnitudes were ≈10−4 N and ≈10−5 N, respectively. It was found that when the radius takes the form of a square root minus a linear function, the profile of the growth force using the models is closer to the experimental force profile. The analysis performed in this work can contribute to a better bubble departure diameter prediction, further enhancing pool boiling analyses with applications in LWR operation.

Abstract The Anticipated Transient Without Scram (ATWS) is classified as a Category II anticipated transient event in nuclear power plants and represents an important subject in reactor safety analysis. Due to the highly integrated configuration of small modular reactors (SMRs), their system responses under accident conditions differ significantly from those of conventional land-based pressurized water reactors. In this study, the thermal-hydraulic system analysis code Relap5 is applied to establish a nodalized model of a representative SMR. On the basis of verified steady-state conditions, a series of transient simulations under design basis accidents (DBAs) combined with ATWS scenarios are performed to evaluate the inherent safety features of the reactor and the accident mitigation capability of its passive safety systems. The results indicate that, for a reactivity insertion accident (RI), the reactor core exhibits strong self-regulating behavior, with power rapidly returning to the steady-state level. In the case of a loss-of-coolant flow accident (LOCF), the large negative feedback coefficients enable the core to reach hot shutdown within a short period, and the decay heat is effectively removed through natural circulation and the passive residual heat removal heat exchangers (PRHRs). For a small break loss of coolant accident (SBLOCA), the passive injection system continuously injects coolant into the core by gravity after the riser section is emptied, ensuring that the upper part of the fuel remains submerged throughout the transient.

ABSTRACT Research on passive cooling for reactor safety has drawn significant attention, with natural convection around confined cylindrical enclosures being a key focus. A computational fluid dynamics (CFD)‐based thermal–hydraulic investigation is conducted to analyze natural convection heat transfer within a 3 × 3 array of vertically heated cylinders enclosed in a water tank. The study employs a full structural detail (FSD) model, accurately representing the geometry of each heating rod to ensure close correspondence with the experimental configuration. Building upon prior single‐phase studies, the present work introduces a two‐phase alumina–water nanofluid model to examine three‐dimensional natural convection phenomena. Numerical simulations are conducted to evaluate the overall Nusselt numbers for both single‐ and symmetric‐cylinder arrangements. Comparison between experimental and CFD predictions confirms strong agreement in the Nusselt–Rayleigh number relationship, validating the numerical methodology.

Probabilistic Safety Assessment (PSA) can be used to evaluate the safety of Nuclear Power Plants (NPPs). To improve NPPs’ safety, reactor safety technology is always developing. Each technology will have a different hazard that can cause an Initiating Event (IE). Of course, this development will affect IE. IE is the event that can affect normal operation and have the potential to cause worst-case conditions when the mitigation system does not work properly. A comprehensive, detailed bibliometric analysis of NPP PSA and IE remains absent from the literature. The objective of this paper is to discuss existing research and identify future trends through bibliometric analysis and a literature review in the PSA and IE domains. The search criteria include the keywords, publication period, publication type, and language. Ninety-five scientific papers were identified during the first screening and were included in the bibliometric analysis. The Bibliometric analysis will represent the keyword network using VOSviewer 1.6.20. The second screening stage is used to strengthen the analysis and generate 43 articles as objects for the literature review. The results showed that the topic of identifying and developing identification methods has not been widely discussed and has become an important research topic. The results of this analysis strengthen the research hypothesis that developing IE methods for new technology NPPs, including High Temperature Gas Cooled Reactors (HTGRs), is feasible.

One of the key concepts in the development of accident-tolerant fuels (ATFs) involves the use of high-density fuels, aimed at improving the safety and economics of light-water reactors. This work focuses on the microstructural and mechanical characterization of experimentally fabricated segments of dispersion-type high-density uranium fuels with a zirconium matrix and Zr-based cladding, intended for future testing in research reactors. Both U-Mo alloys and pure uranium metal were investigated, with particular attention given to the influence of thermal treatment on the phase composition and microstructure of the fuel segments. In addition, changes resulting from thermal exposure were evaluated. The research includes the identification of newly formed phases during transient conditions and the analysis of diffusion phenomena in the material. Mechanical properties were evaluated through microhardness measurements, and the results will serve as input parameters for computational simulations using the Serpent code to support the introduction of these fuels in the VR-2 research reactor.

A safety analysis has been conducted on the RSG-GAS reactor being utilized for 99Mo production from electroplated targets of low-enriched uranium (LEU) positioned in the reactor irradiation position. A static neutronic calculation was performed to investigate the reactivity changes and radial power peaking factor (PPF) while steady-state and transient thermal-hydraulic calculations were carried out to determine the maximum temperatures for the coolant, cladding, and fuel meat. The analysis employed the coupled neutron–thermal-hydraulic code MTRDYN. For optimal 99Mo production, calculations were conducted assuming all central irradiation position and irradiation position facilities were fully loaded with LEU targets. Based on the neutronic calculations, the reactivity changes and maximum radial PPF were found to be 1241.491 pcm and 1.354, respectively. The maximum fuel temperature was 133.35°C for steady-state conditions, while under loss-of-coolant flow, it peaked at 137.94°C. The reactivity-initiated accident transient analysis for positive reactivity insertion reported a maximum coolant temperature of 74.19°C for the 0.0342 $/s reactivity insertion rate, with a cladding temperature of 142.91°C, and a fuel meat maximum temperature of 143.93°C. Both the steady-state and transient calculations showed that the neutronic and thermal-hydraulic parameters were well below the RSG-GAS safety limits for reactor operation, showing the potential for RSG-GAS utilization for 99Mo production from the electroplated LEU target.

Abstract A model‐guided core–shell catalyst design is presented for methanol synthesis, featuring a phase change material (PCM) core encapsulated by a Cu–Zn–AlO (CZA) catalytic shell. The PCM enables in situ thermal management by absorbing reaction heat at its melting point, mitigates the kinetic decline at high temperatures and therefore avoids low conversion, prevents hot spots, and stabilizes the reaction temperature. A two‐dimensional axisymmetric, non‐isothermal packed‐bed reactor model (COMSOL 6.3) was developed for a 10 g system. Simulations evaluate three PCM candidates, that is, LiNO, 9 wt% LiCl + 91 wt% LiNO, and commercial H250, with melting points near 244–250°C. Results indicate that CO conversion can increase from 34.4% to 52.4%, and methanol production can improve by 69% compared to a conventional packed‐bed reactor. Beyond methanol synthesis, the PCM‐integrated core–shell concept provides a scalable approach for thermal control in exothermic reactions, improving reactor efficiency and safety.

The use of passive cooling systems as a reactor safety measure has become a key approach to preventing future incidents similar to the Fukushima Daiichi NPP accident. These systems operate based on natural circulation, a process driven by temperature differences and elevation between the heat source and heat sink. Key design factors, such as the inclination angle of the rectangular loop, significantly influence this circulation. This study aims to investigate the effects of different inclination angles of the rectangular loop and variations in the initial water temperature in the Water Heating Tank (WHT) on the flow rate and heat removal capability. The research was conducted experimentally using a natural circulation rectangular loop facility, FASSIP-04 Ver.0, which has an inner diameter of 26.64 mm, a rectangular loop height of 2280 mm, and a width of 780 mm. The experimental variations were achieved by adjusting the water temperature inside the WHT to 50°C, 70°C, and 90°C. Meanwhile, the inclination angle of the loop was set to 30°, 60°, and 90° mass flow rate and heat removal capability was influenced by both the loop inclination angle and the water temperature in the WHT. The highest values were observed at a 90° inclination angle and a set temperature of 90°C, with a mass flow rate of 0.0241 kg/s, and heat removal rates of qH = 0.791 kW, qC = 0.489 kW. The resulting buoyancy force was stronger under these conditions, leading to greater heat removal through natural circulation compared to free convection, thereby increasing both mass flow rate and heat removal efficiency.

Knowledge of thermal hydraulics is of paramount importance for safety and performance of nuclear reactors. This paper presents the advances in thermal hydraulics in the Dutch Program for Innovation and cOmpetence development for NuclEar infrastructurE and Research (PIONEER). Activities in the fields of system thermal hydraulics, single-phase computational fluid dynamics (CFD), high-resolution CFD, multiphase flow, multiphysics, and fluid-structure interaction are presented illustrated by examples of ongoing work. The main advances in thermal hydraulics within the PIONEER program can be summarized as development and validation of system thermal hydraulics tools as well as pragmatic and high-resolution CFD methodologies and solvers for light water reactors, small modular reactors, and advanced reactor applications. Noteworthy are the developments in the system thermal hydraulics Sophisticated Plant Evaluation Code for Thermal-hydraulic Response Assessment (SPECTRA) tool, multiphase CFD solvers, the generic multiscale coupling tool Multi-phYsics MUltiscale Simulation CoupLing Environment (myMUSCLE), and the open-source multiphysics code platform MUlti-physics Simulation CoupLed Environment based on OpenFOAM (MUSCLE-Foam) as well as the Pressure Fluctuation Model for fluid-structure interaction challenges.