Advances in Thermal Hydraulics of Nuclear Power Plants

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

Deadline for manuscript submissions: 25 July 2025 | Viewed by 1551

Special Issue Editors


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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

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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

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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 (2 papers)

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Research

21 pages, 1634 KiB  
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
Viewed by 266
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 KiB  
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 604
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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