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Thermal Hydraulics and Safety Research for Nuclear Reactors

A special issue of Energies (ISSN 1996-1073). This special issue belongs to the section "B4: Nuclear Energy".

Deadline for manuscript submissions: 25 September 2025 | Viewed by 3885

Special Issue Editors

Dr. Simiao Tang

E-Mail Website
Guest Editor
Key Laboratory of Low-Grade Energy Utilization Technologies and Systems, Ministry of Education, Chongqing University, Chongqing 400044, China
Interests: thermal hydraulics analysis of small nuclear reactor; multi-physics coupling analysis for nuclear reactor; thermal-electrical conversion analysis for nuclear reactor
Dr. Minyang Gui

E-Mail Website
Guest Editor
Heilongjiang Provincial Key Laboratory of Nuclear Power System & Equipment, Harbin Engineering University, Harbin 150001, China
Interests: multi-scale numerical simulation of nuclear reactor; thermal hydraulic characteristics analysis under ocean conditions

Special Issue Information

Dear Colleagues,

The study of key thermal hydraulic phenomena in reactors is of great significance for improving reactor operation safety characteristics and supporting the design and development of next-generation reactors.

This Special Issue examines thermal hydraulics key phenomena and safety research for nuclear reactors and aims to publish papers about thermal hydraulics and safety research technology using a range of experimental, numerical and theoretical methods. The research objects not only include traditional pressurized water reactors, but also involve fourth-generation reactors or other multi-purpose reactors. Accepted papers in this Special Issue will provide important insights into the major issues of current reactor operation and the development of next-generation reactors. By exchanging ideas and shaping the papers for this Special Issue, we will be able to present the most up-to-date technologies for increasing safety characteristics and operating performance of nuclear power plants.

Topics of interest for publication include, but are not limited to, the following:

  1. Thermal hydraulics and safety analysis for nuclear reactors.
  2. Design and engineering advancements in next-generation reactors or small reactors.
  3. Thermal hydraulics key phenomena in next-generation reactors or small reactors.
  4. Design basis accident and severe accident analysis.
  5. Multi-scale/multi-physics simulation for comprehensive reactor analysis.
  6. Modelling, code development and validation.

Dr. Simiao Tang
Dr. Minyang Gui
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 100 words) can be sent to the Editorial Office for announcement on this website.

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. Energies is an international peer-reviewed open access semimonthly 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 2600 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

  • thermal hydraulics
  • safety analysis
  • design and optimization
  • design basis accident
  • severe accident
  • experimental method
  • code development
  • multi-scale/multi-physics simulation
  • PWR
  • next-generation reactor
  • small reactor

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Published Papers (6 papers)

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Research

14 pages, 1622 KiB  
Article
Study on Hydrogen Combustion Flame Acceleration Mechanism and Prediction Method During Severe Accidents in Nuclear Power Plants
by Ran Liu, Jingyi Yu, Xiaoming Yang, Yong Liu, Rubing Ma and Yidan Yuan
Energies 2025, 18(9), 2150; https://doi.org/10.3390/en18092150 - 22 Apr 2025
Viewed by 241
Abstract
Combustion caused by hydrogen-dominated combustible gas mixtures presents critical threats to nuclear safety during severe accidents in nuclear power plants, primarily due to their propensity for flame acceleration, deflagration, and subsequent detonation. Although the direct initiation of detonation from localized hydrogen accumulation at [...] Read more.
Combustion caused by hydrogen-dominated combustible gas mixtures presents critical threats to nuclear safety during severe accidents in nuclear power plants, primarily due to their propensity for flame acceleration, deflagration, and subsequent detonation. Although the direct initiation of detonation from localized hydrogen accumulation at critical concentrations remains challenging, flame acceleration can induce rapid pressure escalation and lead to deflagration-to-detonation transition under suitable conditions. The ultra-high-pressure loads generated almost instantaneously will pose serious risks to containment integrity and equipment or instrument functionality within nuclear facilities. This paper investigates the flame acceleration mechanism and criterion, which is crucial for precise hydrogen risk assessment. A hydrogen combustion flame acceleration model is developed, integrating both laminar and turbulent flame propagation across multiple control volumes. Validated against the RUT test, the model demonstrates high fidelity with a maximum 3.17% deviation in flame propagation velocity and successfully captures the pressure discontinuity. The developed model enables comprehensive simulation with improved predictive accuracy of the flame acceleration process, making it an essential tool for advancing fundamental understanding of hydrogen behavior and severe accident analysis. This model’s development marks a paradigm in nuclear safety research by providing an analytical instrument for integrated severe accident programs in nuclear power plants, contributing to improving the potential hydrogen risks assessment and management in next-generation reactor safety. Full article
(This article belongs to the Special Issue Thermal Hydraulics and Safety Research for Nuclear Reactors)
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24 pages, 16076 KiB  
Article
Instability Analysis of Two-Phase Flow in Parallel Rectangular Channels for Compact Nuclear Reactors
by Simiao Tang, Can Wang, Zaiyong Ma, Calvin Febianto Liem, Quanyao Ren, Qiang Lian, Longxiang Zhu, Luteng Zhang, Wan Sun, Meiyue Yan and Liangming Pan
Energies 2025, 18(8), 2049; https://doi.org/10.3390/en18082049 - 16 Apr 2025
Viewed by 266
Abstract
In this paper, a numerical study of two-phase flow instability in parallel rectangular channels is presented. Using the homogeneous flow model, marginal stability boundaries (MSBs) are derived in the parameter space defined by the phase change number (Npch) and subcooling number [...] Read more.
In this paper, a numerical study of two-phase flow instability in parallel rectangular channels is presented. Using the homogeneous flow model, marginal stability boundaries (MSBs) are derived in the parameter space defined by the phase change number (Npch) and subcooling number (Nsub) under various operating conditions. Comparison with experimental data shows that the model predicts stability trends with a deviation of ±12.5%. The study reveals that, under constant mass flux conditions, stability decreases as the equivalent diameter (De) of the channels increases. Additionally, the exit area ratio of the two parallel tubes has minimal effect on the MSB, indicating that exit geometry does not significantly influence system stability. However, an increase in the inlet area ratio, from 0.1 to 1, reduces system stability, suggesting that larger inlet areas relative to tube cross-sectional areas may lead to greater flow disturbances, thereby decreasing stability. Moreover, increasing the length of the tubes enhances system stability, which may be attributed to the extended development length allowing for dissipation of flow disturbances. The study further demonstrates that higher flow rates, between 0.15 kg/s and 0.25 kg/s, enhance stability, while increasing the outlet flow resistance coefficient reduces stability. Conversely, increasing the inlet flow resistance coefficient improves stability. At system pressures of 3 MPa, 6 MPa, and 9 MPa, it is observed that higher pressures shift the boundary of complete vaporization (Xe = 1) to the left on the Npch and Nsub graph, reducing the region susceptible to instability. The study also employs Fast Fourier Transform (FFT) analysis to identify peak frequencies across different parameter ranges. By examining the stability map and frequency spectra, the study provides deeper insights into two-phase flow instabilities in parallel channels. Full article
(This article belongs to the Special Issue Thermal Hydraulics and Safety Research for Nuclear Reactors)
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13 pages, 4746 KiB  
Article
Research on the Impact of Heating Conditions for Passive Air-Cooling System Wind Loading Performance Test
by Hongliang Wang, Mingrui Yu, Yong Guo, Yu Feng, Wei Li, Qiang Guo and Yidan Yuan
Energies 2025, 18(7), 1670; https://doi.org/10.3390/en18071670 - 27 Mar 2025
Viewed by 137
Abstract
The wind loading test serves as a critical validation experiment for the Passive Air-Cooling System (PAS) ACP100. However, it remains unclear how the highly scaled-down experimental setup can accurately account for the influence of the steel shell wall heating conditions on the airflow [...] Read more.
The wind loading test serves as a critical validation experiment for the Passive Air-Cooling System (PAS) ACP100. However, it remains unclear how the highly scaled-down experimental setup can accurately account for the influence of the steel shell wall heating conditions on the airflow dynamics within the PAS flow channel. This study employs both numerical simulations and experimental investigations to compare and analyze variations in pressure and temperature within the PAS flow channel under the different heating temperatures of a steel shell wall, considering scenarios with and without environmental wind field effects. The objective is to assess the influence and necessity of heating conditions. In this study, ANSYS Fluent 18.2 was utilized to conduct numerical simulations of the 1:126 scale model of ACP100. Subsequently, the 1:126-scale ACP100 test model was placed on a wind tunnel platform to investigate various experimental conditions. Key parameters, including pressure, temperature, and wind velocity, were meticulously measured at critical locations to obtain detailed insights into the model’s performance under different scenarios. The results indicate that the numerical calculations are consistent with the findings from experimental research. When the environmental wind velocity is 0 m/s, the pressure deviation (∆Pmax) at each measurement position within the PAS flow channel, under varying heating wall temperatures of 55.8 °C, 93.5 °C and 126.8 °C, remains below 1.8 Pa. Furthermore, the inlet and outlet pressure difference (∆Pio) is less than 3.9 Pa, which is insufficient to establish natural circulation. Additionally, it was observed that the air temperature increases continuously from the PAS inlet to the top outlet; notably, the air temperature at the top outlet approaches that of the heating wall temperature, nearly reaching equilibrium. When examining the coupling effect of the environmental wind field, it is observed that the pressure difference (∆Pio) between the inlet and top outlet of the PAS flow channel increases significantly. However, the pressure deviation at each measurement position within the PAS flow channel remains within acceptable limits, satisfying ∆Pmax < 3 Pa. Furthermore, the temperature deviation (∆Tmax ≤ 2.8 °C) at each measuring position in the PAS channel indicates that the influence of the environmental wind field on both pressure and temperature distribution is relatively minor and can be safely neglected. In summary, it can be concluded that when utilizing the ACP100 scaled-down model for research on a PAS wind loading performance test, there is no necessity to establish heating conditions as their effects can be disregarded. Full article
(This article belongs to the Special Issue Thermal Hydraulics and Safety Research for Nuclear Reactors)
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23 pages, 14284 KiB  
Article
Development and Performance Analysis of an Electromagnetic Pump for a Thermal Hydraulic Experimental Loop of a Lead-Cooled Fast Reactor
by Zi’ang Li, Lanfei Yuan, Chenglong Wang, Suizheng Qiu and Ying Li
Energies 2025, 18(3), 750; https://doi.org/10.3390/en18030750 - 6 Feb 2025
Viewed by 769
Abstract
With the advancement of lead–bismuth fast reactors, there has been increasing attention directed towards the design of and manufacturing technology for electromagnetic pumps employed to drive liquid lead–bismuth eutectic (LBE). These electromagnetic pumps are characterized by a simple structure, effective sealing, and ease [...] Read more.
With the advancement of lead–bismuth fast reactors, there has been increasing attention directed towards the design of and manufacturing technology for electromagnetic pumps employed to drive liquid lead–bismuth eutectic (LBE). These electromagnetic pumps are characterized by a simple structure, effective sealing, and ease of flow control. They exploit the excellent electrical conductivity of liquid metals, allowing the liquid metal to be propelled by Lorentz forces generated by the traveling magnetic field within the pump. To better understand the performance characteristics of electromagnetic pumps and master the techniques for integrated manufacturing and performance optimization, this study conducted fundamental research, development of key components, and the assembly of the complete pump. Consequently, an annular linear induction pump (ALIP) suitable for liquid lead–bismuth eutectic was developed. Additionally, within the lead–bismuth thermal experimental loop, startup and preheating experiments, performance tests, and flow-head experiments were conducted on this electromagnetic pump. The experimental results demonstrated that the output flow of the electromagnetic pump increased linearly with the input current. When the input current reached 99 A, the loop achieved a maximum flow rate of 8 m3/h. The efficiency of the electromagnetic pump also increased with the input current, with a maximum efficiency of 5.96% during the experiments. Finally, by analyzing the relationship between the flow rate and the pressure difference of the electromagnetic pump, a flow-head model specifically applicable to lead–bismuth electromagnetic pumps was established. Full article
(This article belongs to the Special Issue Thermal Hydraulics and Safety Research for Nuclear Reactors)
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24 pages, 10275 KiB  
Article
New Nusselt Number Correlation and Turbulent Prandtl Number Model for Turbulent Convection with Liquid Metal Based on Quasi-DNS Results
by Hao Fu, Juan Chen, Yanjun Tong, Sifan Peng, Fang Liu, Xuefeng Lyu and Houjian Zhao
Energies 2025, 18(3), 547; https://doi.org/10.3390/en18030547 - 24 Jan 2025
Viewed by 973
Abstract
Liquid metal is widely used as the primary coolant in many advanced nuclear energy systems. Prandtl number of liquid metal is much lower than that of the conventional coolant of water or gas. Based on the Reynolds analogy, the turbulent Prandtl number is [...] Read more.
Liquid metal is widely used as the primary coolant in many advanced nuclear energy systems. Prandtl number of liquid metal is much lower than that of the conventional coolant of water or gas. Based on the Reynolds analogy, the turbulent Prandtl number is assumed to be a constant around unity. For the turbulent convection of liquid metal, dissipations of half the temperature variance are larger than those of turbulent kinetic energies. The dissimilarity between the thermal and momentum fields increases as Pr decreases. The turbulent Prandtl number is larger than one for the liquid metal. In the current investigation, the turbulent convection of liquid metal in the channel is quasi-directly simulated with OpenFOAM-7. The turbulent statistics of the momentum and the thermal field are compared with the existing database to validate the numerical model. The power law for dimensionless temperature distribution with different Prandtl numbers is obtained by regression analysis of numerical results. A new Nusselt number correlation is derived based on the power law. The new Nusselt number correlation agrees well with the DNS results in the literature. The momentum mixing process between different layers in the cross section is compared with the thermal mixing process. The effects of the Prandtl number on the difference between the turbulence time scale and scalar time scale are analyzed. A new turbulent Prandtl number model with local parameters is obtained for turbulent convection with liquid metal. Combined with the k−ω model, the temperature distributions with the new turbulent Prandtl number model agree well with the DNS results in the literature. The new turbulent Prandtl number model can be used for turbulent convection with different Prandtl and different Reynolds numbers. Full article
(This article belongs to the Special Issue Thermal Hydraulics and Safety Research for Nuclear Reactors)
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23 pages, 10946 KiB  
Article
Dynamic Multiphysics Simulation of the Load-Following Behavior in a Typical Pressurized Water Reactor Power Plant
by Ivan Panciak and Aya Diab
Energies 2024, 17(24), 6373; https://doi.org/10.3390/en17246373 - 18 Dec 2024
Cited by 1 | Viewed by 847
Abstract
Most Nuclear Power Plants (NPPs) are designed for baseload operations, maintaining a steady power output at 100%, except during planned maintenance and refueling. However, in countries like France, Slovakia, and Korea, where nuclear power is a major source of electricity, integrating nuclear energy [...] Read more.
Most Nuclear Power Plants (NPPs) are designed for baseload operations, maintaining a steady power output at 100%, except during planned maintenance and refueling. However, in countries like France, Slovakia, and Korea, where nuclear power is a major source of electricity, integrating nuclear energy with intermittent renewables is crucial for stable power generation. This integration necessitates daily power adjustments by NPPs in response to grid demands, a process known as a Load Follow Operation (LFO). Such a process introduces strong interdependencies between thermal–hydraulic and neutron–kinetic parameters, coupled with the three-dimensional movement of Control Element Assemblies (CEAs) and Xenon dynamics, which pose safety challenges due to shifts in core power distribution. To address these complexities, a multi-physics approach is employed using the multi-physics package RELAP5/3DKIN and implementing two strategies. The first strategy uses a mechanical shim, adjusting the reactor power exclusively through CEAs. The second strategy combines CEA movement with adjustments in soluble boron concentration. Both strategies are evaluated against axial offset and 3D power peaking safety limits to ensure compliance with operational safety requirements. Full article
(This article belongs to the Special Issue Thermal Hydraulics and Safety Research for Nuclear Reactors)
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