Advances in Nuclear Reactor Thermal-Hydraulics: From Fundamental Research to Advanced Reactor Applications

This Research Topic is associated with the 6th Annual Conference of the CNS Nuclear Reactor Thermal-Hydraulics Division. We welcome submissions from attendees at the 6th Annual Conference of the CNS Nuclear Reactor Thermal-Hydraulics Division and others conducting relevant research.Nuclear reactor thermal-hydraulics is central to the safety, performance, economic viability, and licensing of nuclear energy systems, supporting reactor design, operation, accident analysis, and the development of emerging concepts. The expansion of water-cooled reactors, small modular reactors (SMRs), Generation IV systems, microreactors, and fusion-related technologies is creating demanding research questions involving multi-phase flow, boiling, critical heat flux, phase change, turbulence, and multi-scale, multi-physics interactions. Recent advances in high-fidelity computational fluid dynamics, experimental diagnostics, system and sub-channel codes, uncertainty quantification, and data-driven modelling are improving predictive capabilities. However, challenges remain in model closure, scale bridging, validation under representative conditions, extrapolation to novel reactor environments, and the reliable integration of physical and machine-learning models. Further investigation is needed to connect fundamental understanding with robust engineering solutions and to support safety assessment, accident management, and the transition from empirical design practices toward physics-informed and data-enabled methodologies.This Research Topic aims to present advances and identify key challenges in nuclear reactor thermal-hydraulics across fundamental research, computational methods, experimentation, reactor engineering, and safety analysis. It seeks to deepen understanding of flow and heat-transfer phenomena; promote the development, validation, and application of high-fidelity, multi-scale, and machine-learning-enhanced simulations; advance measurement, online analysis, and uncertainty-quantification techniques; and address thermal-hydraulic requirements in water-cooled, advanced, small modular, and other emerging reactor systems. Contributions are also encouraged on severe-accident phenomena and mitigation strategies, with particular attention to improving predictive reliability, supporting code verification and validation, and strengthening the technical basis for nuclear safety and regulatory assessment.To gather further insights across fundamental reactor thermal-hydraulics, computational methods, experimental technologies, reactor applications, and safety analysis, this Research Topic welcomes Original Research, Review, Methods, Technology and Code, Brief Research Report, and Perspective articles. We encourage submissions addressing, but not limited to, the following themes:• Fundamental Theories and Phenomena in Reactor Thermal-Hydraulics: Mechanisms of single-phase and two-phase flow and heat transfer; critical heat flux (CHF); flow instability; multi-scale and multi-physics coupling theories.• Computational Fluid Dynamics (CFD) in Nuclear Reactors: Turbulence modeling; interface tracking; boiling and heat transfer modeling; conjugate heat transfer (CHT); advanced numerical methods and high-performance computing (HPC) applications.• Thermal-Hydraulic Experiments and Advanced Measurement Techniques: Test facility design; advanced diagnostics (optical diagnostics, ultrasound, neutron radiography, etc.); experimental data processing and uncertainty quantification.• Thermal-Hydraulics in Water-Cooled Reactor Design and Operation: Core thermal-hydraulic design; fuel assembly thermal performance; operational transients and load following; component aging and plant life management.• Thermal-Hydraulics in Advanced and Innovative Reactor Concepts: Small Modular Reactors (SMRs); Generation IV advanced reactors (sodium-cooled, lead-cooled, gas-cooled, and molten salt reactors); microreactors; and thermal-hydraulic studies supporting nuclear fusion systems.• Severe Accidents and Mitigation Strategies in Nuclear Power Plants: Severe accident phenomenology; core degradation, corium behavior, and molten pool dynamics; hydrogen risk and mitigation; containment thermal-hydraulics; accident management guidelines (SAMGs).• Code Development, Simulation, and Validation & Verification (V&V): System code and sub-channel code development; Best Estimate Plus Uncertainty (BEPU) methodologies; code V&V; system-level and component-level numerical simulations.• Interdisciplinary and Emerging Frontiers in Reactor Thermal-Hydraulics: Cross-cutting research integrating thermal-hydraulics with materials science, structural mechanics, control systems, and artificial intelligence/machine learning (AI/ML), as well as other emerging topics in the field.Topic Editors, Long Gu, Wenxing Liu, and Mingjun Wang declare that they are involved in the organization of the 6th Annual Conference of the CNS Nuclear Reactor Thermal-Hydraulics Division. Topic editors: Mingjun Wang, Wenxing LIU, Long Gu Source: https://www.frontiersin.org/research-topics/86435/advances-in-nuclear-reactor-thermal-hydraulics-from-fundamental-research-to-advanced-reactor-applications