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Advances in Thermal Hydraulics of Nuclear Power Plants
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Advances in Thermal Hydraulics of Nuclear Power Plants

A special issue of Journal of Nuclear Engineering (ISSN 2673-4362).

Deadline for manuscript submissions: closed (31 December 2025) | Viewed by 14601

Special Issue Editors

Dr. Milica Ilic

E-Mail Website
Guest Editor
Vinca Institute of Nuclear Sciences, National Institute of the Republic of Serbia, University of Belgrade, Belgrade, Serbia
Interests: mechanical engineering; thermal-hydraulics of fusion and nuclear reactors; design of small modular reactors; two-phase flow phenomena in power utilities; energy storage and flexibility of power plants
Dr. Piyush Sabharwall

E-Mail Website
Guest Editor
Reactor Systems Design and Analysis Division, Nuclear Science & Technology, Idaho National Laboratory, Idaho Falls, ID 83415, USA
Interests: thermal hydraulics; advanced nuclear reactors; nuclear safety systems; energy systems; heat transfer and fluid mechanics; integrated energy systems; experiments and modeling; heat exchangers; system and component modeling

Special Issue Information

Dear Colleagues,

From its beginning, nuclear power engineering has relied heavily on thermal hydraulics. The thermal hydraulics systems were, and have remained, the essence of safe operation of nuclear power plants during steady and transient regimes as well as in accident situations. The investigations of thermal hydraulics phenomena have been expanding to involve more and more convoluted fluid flow and heat transfer mechanisms under conditions of high heat fluxes, within complex flow domains and for diverse coolants.

This Special Issue, titled “Advances in Thermal Hydraulics of Nuclear Power Plants”, of JNE is devoted to thermal hydraulics in a contemporary fleet of nuclear power plants and in reactor technologies under development. This Special Issue is thought to be a collection of research papers in which both the phenomena and the applied technical solutions in the field of nuclear thermal hydraulics are addressed through comprehensive discussions, modeling, numerical simulations and experimentation. The authors are invited to submit manuscripts which deal with thermal hydraulics research in light water PWR and BWR reactors, CANDU reactors, liquid metal reactors, molten salt reactors, high temperature gas reactors, fast reactors and other advanced nuclear reactors, small modular reactors and micro-reactors, as well as in fusion reactors. The Special Issue welcomes manuscripts which deal with thermal hydraulics mechanisms such as two-phase flow, phase transition, critical heat flux and dryout, severe accidents (LOCA, core melting and subsequent debris bed cooling), flow in subchannels with rod bundles, heat transfer in steam generators, passive cooling systems, thermal hydraulics of containment and hydrogen generation, extraction of heat from plasma facing components, managing high heat fluxes in divertor and thermal hydraulics of other fusion reactor components. Review papers on thermal hydraulics in nuclear power plants in general or in a specific field will be appreciated.

Dr. Milica Ilic
Dr. Piyush Sabharwall
Guest Editors

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All submissions that pass pre-check are peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 250 words) can be sent to the Editorial Office for assessment.

Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Journal of Nuclear Engineering is an international peer-reviewed open access quarterly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 1200 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.

Keywords

  • computational fluid dynamics
  • thermal hydraulics system codes
  • thermal hydraulics experiments
  • advanced and Gen IV reactors
  • small modular reactors
  • fusion reactors
  • passive systems
  • transient and severe accident conditions
  • multi-phase flow and phase transition
  • containment thermal hydraulics

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Published Papers (9 papers)

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Research

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34 pages, 13512 KB  
Article
Performance and Scalability Analysis of Hydrodynamic Fluoride Salt Lubricated Bearings in Fluoride-Salt-Cooled High-Temperature Reactors
by Yuqi Liu and Minghui Chen
J. Nucl. Eng. 2026, 7(1), 11; https://doi.org/10.3390/jne7010011 - 29 Jan 2026
Viewed by 233
Abstract
This study evaluates the performance and scalability of fluoride-salt-lubricated hydrodynamic journal bearings used in primary pumps for Fluoride-salt-cooled High-temperature Reactors (FHRs). Because full-scale pump prototypes have not been tested, a scaling analysis is used to relate laboratory results to commercial conditions. Bearings with [...] Read more.
This study evaluates the performance and scalability of fluoride-salt-lubricated hydrodynamic journal bearings used in primary pumps for Fluoride-salt-cooled High-temperature Reactors (FHRs). Because full-scale pump prototypes have not been tested, a scaling analysis is used to relate laboratory results to commercial conditions. Bearings with different length-to-diameter (L/D) ratios were assessed over a range of shaft speeds to quantify geometric and hydrodynamic effects. High-temperature bushing test data in FLiBe at 650 °C were used as inputs to three-dimensional computational fluid dynamics (CFD) simulations in STAR-CCM+. Applied load, friction force, and power loss were computed across operating speeds. Applied load increases linearly with shaft speed due to hydrodynamic pressure buildup, while power loss increases approximately quadratically, indicating greater energy dissipation at higher speeds. The resulting correlations clarify scaling effects beyond small-scale testing and provide a basis for bearing design optimization, prototype development, and the deployment of FHR technology. This work benchmarks speed-scaling relations for fluoride-salt-lubricated hydrodynamic journal bearings within the investigated regime. Full article
(This article belongs to the Special Issue Advances in Thermal Hydraulics of Nuclear Power Plants)
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15 pages, 5972 KB  
Article
Thermal Hydraulics and Solid Mechanics Multiphysics Safety Analysis of a Heavy Water Reactor with Thorium-Based Fuel
by Bayan Kurbanova, Yuriy Sizyuk, Ansar Aryngazin, Zhanna Alsar, Ahmed Hassanein and Zinetula Insepov
J. Nucl. Eng. 2025, 6(4), 53; https://doi.org/10.3390/jne6040053 - 30 Nov 2025
Viewed by 723
Abstract
Growing environmental awareness has renewed interest in thorium as a nuclear fuel, underscoring the need for further studies to evaluate how reactors perform when conventional fuels are replaced with thorium-based alternatives. In this study, thermal hydraulics and solid mechanics computations were simulated using [...] Read more.
Growing environmental awareness has renewed interest in thorium as a nuclear fuel, underscoring the need for further studies to evaluate how reactors perform when conventional fuels are replaced with thorium-based alternatives. In this study, thermal hydraulics and solid mechanics computations were simulated using COMSOL multiphysics to investigate the safe operating conditions of a heavy water reactor with thorium-based fuel. The thermo-mechanical analysis of the fuel rod under transient heating conditions provides critical insights into strain, displacement, stress, and coolant flow behavior at elevated volumetric heat sources. After 3 s of heating, the strain distribution in the fuel exhibits a high-strain core surrounded by a low-strain rim, with peak volumetric strain increasing nearly linearly from 0.006 to 0.014 as heat generation rises. Displacement profiles confirm that radial deformation is concentrated at the outer surface, while axial elongation remains uniform and scales systematically with power. The resulting von Mises stress fields show maxima at the outer surface, increasing from ~0.06 to 0.15 GPa at the centerline with higher heat input but remaining within structural safety margins. Cladding simulations demonstrate nearly uniform axial expansion, with displacements increasing from ~0.012 mm to 0.03 mm across the investigated power range, and average strain remains negligible (≈10−4), while mean stresses increase moderately yet stay well below the yield strength of zirconium alloys, confirming safe elastic behavior. Hydrodynamic analysis shows that coolant velocity decreases smoothly along the axial direction but maintains stability, with only minor reductions under increased heat sources. Overall, the coupled thermo-mechanical and fluid-dynamic results confirm that both the fuel and cladding remain structurally stable under the studied conditions. By using COMSOL’s multiphysics capabilities, and unlike most legacy codes optimized for uranium-based fuel, this work is designed to easily incorporate non-traditional fuels such as thorium-based systems, including user-defined material properties, temperature-dependent thermal polynomial formulas, and mechanical response. Full article
(This article belongs to the Special Issue Advances in Thermal Hydraulics of Nuclear Power Plants)
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15 pages, 3928 KB  
Article
Numerical Investigation of Fluid–Structure Interaction of Foreign Objects in Steam Generator Tube Bundles
by Yuhua Hang, Heng Wang, Yuanqing Liu, Zhen Cai, Bin Zhu, Jinna Mei and Guorui Zhu
J. Nucl. Eng. 2025, 6(4), 47; https://doi.org/10.3390/jne6040047 - 19 Nov 2025
Viewed by 618
Abstract
As a critical component of nuclear and thermal energy conversion systems, the long-term safe operation of a steam generator depends on the structural integrity of its tube bundles. Foreign objects introduced into the secondary side can induce flow-induced vibrations and wear, potentially causing [...] Read more.
As a critical component of nuclear and thermal energy conversion systems, the long-term safe operation of a steam generator depends on the structural integrity of its tube bundles. Foreign objects introduced into the secondary side can induce flow-induced vibrations and wear, potentially causing tube wall damage and unplanned outages, thereby affecting overall system reliability. This study systematically investigates the flow-induced vibration behavior of foreign objects within steam generator tube bundles and explores the influence of object geometry through three-dimensional fluid–structure interaction (FSI) simulations. The foreign objects are modeled as single-degree-of-freedom rigid bodies, and their dynamic responses are captured using a coupled flow–motion framework. Results reveal that object geometry significantly influences flow separation, variations in lift and drag forces, and displacement characteristics. Cylindrical and irregular objects exhibit stable, low-amplitude vibrations; plate-shaped objects experience restricted motion due to large drag areas and symmetric contact constraints; whereas helical objects show the largest displacements arising from coupled axial–radial vibrations and complex vortical structures. These findings demonstrate that the interplay between aerodynamic forces and geometric complexity strongly governs the flow-induced vibration of foreign objects, offering insights into their motion behavior and potential impact on steam generator tube bundle integrity. Full article
(This article belongs to the Special Issue Advances in Thermal Hydraulics of Nuclear Power Plants)
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22 pages, 9259 KB  
Article
Two-Phase Flow Studies in Steam Separators Using Interface Capturing Simulations
by Taylor E. Grubbs and Igor A. Bolotnov
J. Nucl. Eng. 2025, 6(4), 42; https://doi.org/10.3390/jne6040042 - 15 Oct 2025
Viewed by 1150
Abstract
The two-phase flow within a Boiling Water Reactor steam separator is investigated using an interface capturing method. The simulations are focused on resolving the flow around the first pickoff ring which is the highest contributor to steam carryunder phenomenon. Multiple simulations are conducted [...] Read more.
The two-phase flow within a Boiling Water Reactor steam separator is investigated using an interface capturing method. The simulations are focused on resolving the flow around the first pickoff ring which is the highest contributor to steam carryunder phenomenon. Multiple simulations are conducted of varying levels of resolution to evaluate the capabilities of interface capturing technique for this challenging problem. First, high-resolution simulations of the flow using a simplified 30° wedge are conducted without a swirling velocity field present in the actual system. In order to understand the flow field generated by the separator swirler, secondary simulations of single-phase flow passing through a swirler model are conducted. Using this information, a coarse simulation of the full 360° model was performed, which incorporated the effect of the swirler using a custom inflow boundary condition. Instantaneous carryunder/carryover along with void fraction and film thickness are evaluated at the pickoff ring entrance. Overall, these simulations demonstrate that interface capturing simulations can be an accurate tool for studying full-scale components within nuclear power plants. Full article
(This article belongs to the Special Issue Advances in Thermal Hydraulics of Nuclear Power Plants)
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26 pages, 6112 KB  
Article
Preliminary Experimental Validation of Single-Phase Natural Circulation Loop Based on RELAP5-3D Code: Part I
by Hossam H. Abdellatif, Joshua Young, David Arcilesi and Richard Christensen
J. Nucl. Eng. 2025, 6(3), 38; https://doi.org/10.3390/jne6030038 - 19 Sep 2025
Viewed by 1735
Abstract
The molten salt reactor (MSR) is a prominent Generation IV nuclear reactor concept that offers substantial advantages over conventional solid-fueled systems, including enhanced fuel utilization, inherent passive safety features, and significant reductions in long-lived radioactive waste. Central to its safety strategy is a [...] Read more.
The molten salt reactor (MSR) is a prominent Generation IV nuclear reactor concept that offers substantial advantages over conventional solid-fueled systems, including enhanced fuel utilization, inherent passive safety features, and significant reductions in long-lived radioactive waste. Central to its safety strategy is a reliance on natural circulation (NC) mechanisms, which eliminate the need for active pumping systems and enhance system reliability during normal and off-normal conditions. However, the challenges associated with molten salts, such as their high melting points, corrosivity, and material compatibility issues, render experimental investigations inherently complex and demanding. Therefore, the use of high-Pr-number surrogate fluids represents a practical alternative for studying molten salt behavior under safer and more accessible experimental conditions. In this study, a single-phase natural circulation loop setup at the University of Idaho’s Thermal–Hydraulics Laboratory was employed to investigate NC behavior under various operating conditions. The RELAP5-3D code was initially validated against water-based experiments before employing Therminol-66, a high-Prandtl-number surrogate for molten salts, in the natural circulation loop for the first time. The RELAP5-3D results demonstrated good agreement with both steady-state and transient experimental results, thereby confirming the code’s ability to model NC behavior in a single-phase flow regime. The results also highlighted certain experimental limitations that should be addressed to enhance the NC loop’s performance. These include increasing the insulation thickness to reduce heat losses, incorporating a dedicated mass flow measurement device for improved accuracy, and replacing the current heater with a higher-capacity unit to enable testing at elevated power levels. By identifying and addressing the main causes of these limitations and uncertainties during water-based experiments, targeted improvements can be implemented in both the RELAP5 model and the experimental setup, thereby ensuring that tests using a surrogate fluid for MSR analyses are conducted with higher accuracy and minimal uncertainty. Full article
(This article belongs to the Special Issue Advances in Thermal Hydraulics of Nuclear Power Plants)
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18 pages, 2791 KB  
Article
Deterministic Data Assimilation in Thermal-Hydraulic Analysis: Application to Natural Circulation Loops
by Lanxin Gong, Changhong Peng and Qingyu Huang
J. Nucl. Eng. 2025, 6(3), 23; https://doi.org/10.3390/jne6030023 - 3 Jul 2025
Viewed by 1248
Abstract
Recent advances in high-fidelity modeling, numerical computing, and data science have spurred interest in model-data integration for nuclear reactor applications. While machine learning often prioritizes data-driven predictions, this study focuses on data assimilation (DA) to synergize physical models with measured data, aiming to [...] Read more.
Recent advances in high-fidelity modeling, numerical computing, and data science have spurred interest in model-data integration for nuclear reactor applications. While machine learning often prioritizes data-driven predictions, this study focuses on data assimilation (DA) to synergize physical models with measured data, aiming to enhance predictive accuracy and reduce uncertainties. We implemented deterministic DA methods—Kalman filter (KF) and three-dimensional variational (3D-VAR)—in a one-dimensional single-phase natural circulation loop and extended 3D-VAR to RELAP5, a system code for two-phase loop analysis. Unlike surrogate-based or model-reduction strategies, our approach leverages full-model propagation without relying on computationally intensive sampling. The results demonstrate that KF and 3D-VAR exhibit robustness against varied noise types, intensities, and distributions, achieving significant uncertainty reduction in state variables and parameter estimation. The framework’s adaptability is further validated under oceanic conditions, suggesting its potential to augment baseline models beyond conventional extrapolation boundaries. These findings highlight DA’s capacity to improve model calibration, safety margin quantification, and reactor field reconstruction. By integrating high-fidelity simulations with real-world data corrections, the study establishes a scalable pathway to enhance the reliability of nuclear system predictions, emphasizing DA’s role in bridging theoretical models and operational demands without compromising computational efficiency. Full article
(This article belongs to the Special Issue Advances in Thermal Hydraulics of Nuclear Power Plants)
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21 pages, 1634 KB  
Article
Droplet Entrainment in Steam Supply System of Water-Cooled Small Modular Reactors: Experiment and Modeling Approaches
by Kenneth Lee Fossum, Palash Kumar Bhowmik and Piyush Sabharwall
J. Nucl. Eng. 2024, 5(4), 563-583; https://doi.org/10.3390/jne5040035 - 12 Dec 2024
Cited by 3 | Viewed by 2690
Abstract
Droplet entrainment in steam-flow is a prominent phenomenon that needs adequate safety and risk analysis of postulated transient and accident scenarios—including experimental investigation and representative modeling and simulation (M&S)—for small modular reactor (SMR) system design and demonstration. This study identifies knowledge gaps by [...] Read more.
Droplet entrainment in steam-flow is a prominent phenomenon that needs adequate safety and risk analysis of postulated transient and accident scenarios—including experimental investigation and representative modeling and simulation (M&S)—for small modular reactor (SMR) system design and demonstration. This study identifies knowledge gaps by evaluating experimental and computational fluid dynamics modeling approaches to support early-stage reactor system design, testing, and model evaluation. Previous studies reported in the literature for steam-flow entrainment primarily focused on gigawatt capacity pressurized water reactor (PWR) systems. However, entrainment phenomena are even more prominent for PWR-type SMRs due to their more compact integrated designs, which need further research and development. To fill the research gaps, this study provides insight by specifying the phenomena of interest by leveraging the lessons learned from past research, adopting advanced M&S techniques and advanced instrumentation and control. The findings and recommendations are applicable for evaluating steam-flow entrainment models and for designing integral effect test and separate effect test facilities for gaining reactor design approvals. Full article
(This article belongs to the Special Issue Advances in Thermal Hydraulics of Nuclear Power Plants)
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11 pages, 5542 KB  
Article
Experimental and Numerical Study on the Characteristics of Bubble Motion in a Narrow Channel
by Borong Tang, Shenfei Wang, Fang Liu and Fenglei Niu
J. Nucl. Eng. 2024, 5(4), 445-455; https://doi.org/10.3390/jne5040028 - 15 Oct 2024
Viewed by 2126
Abstract
Plate fuel elements, known for their compact structure and efficient cooling, are commonly used in the core of nuclear reactors. In these reactors, coolant channels are designed as rectangular narrow slits. Bubble behavior in narrow channels differs significantly from that in conventional channels. [...] Read more.
Plate fuel elements, known for their compact structure and efficient cooling, are commonly used in the core of nuclear reactors. In these reactors, coolant channels are designed as rectangular narrow slits. Bubble behavior in narrow channels differs significantly from that in conventional channels. This paper investigates the vertical rise of bubbles in narrow slit channels. A gas–liquid two-phase flow experimental rig was constructed using transparent acrylic boards. A high-speed camera captured the bubble formation process during gas injection, and code implemented in Matlab was used to process the images. Numerical simulations were conducted with CFD software under identical conditions and compared with the experimental results, showing a good agreement. The results show that the experimental and simulated bubble movement velocities are in good agreement. In the experiments of this paper, when the width of the narrow gap is below 3 mm, the sidewalls exert a pronounced influence on the dynamics of bubble rise, notably altering both the velocity profile and the trajectory of the bubbles’ ascent. As the gas injection flow rate gradually increases, the bubble rising speed and trajectory change from regular to oscillatory patterns. Full article
(This article belongs to the Special Issue Advances in Thermal Hydraulics of Nuclear Power Plants)
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Review

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29 pages, 14985 KB  
Review
Towards a Universal System for the Classification of Boiling Surfaces
by Alexander Ustinov, Jovan Mitrovic and Dmitry Ustinov
J. Nucl. Eng. 2025, 6(1), 7; https://doi.org/10.3390/jne6010007 - 12 Mar 2025
Cited by 2 | Viewed by 1627
Abstract
A lot of novel surface treatment technologies have appeared over the last few decades, offering great possibilities for practical use. Modified surfaces have confirmed their successful application in thermal engineering for boiling heat transfer enhancement and single-phase convection. Several classification approaches for boiling [...] Read more.
A lot of novel surface treatment technologies have appeared over the last few decades, offering great possibilities for practical use. Modified surfaces have confirmed their successful application in thermal engineering for boiling heat transfer enhancement and single-phase convection. Several classification approaches for boiling surfaces exist in the literature; however, a full, physically based, and commonly accepted universal system is still missing. This paper proposes such a classification system, based on considerations of physical mechanisms underlying the nucleation process and enhancement mechanism during different stages of vapor bubble growth. It also presents an overview of recent advances in the development of enhanced boiling surfaces. Full article
(This article belongs to the Special Issue Advances in Thermal Hydraulics of Nuclear Power Plants)
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