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Special Issue "Thermal-Hydraulic Challenges in Advanced Nuclear Reactors"

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

Deadline for manuscript submissions: closed (30 September 2023) | Viewed by 8145

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Special Issue Editor

Prof. Dr. Naihua Wang

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Guest Editor

Institute of Thermal Science and Technology, Shandong University, Jinan 250061, China
Interests: nuclear thermal hydraulic; thermal management; heat and mass transfer; two-phase flow

Special Issue Information

Dear Colleagues,

Nuclear engineering and technology play a vital role in achieving low carbon emission goals worldwide, while providing reliable, baseload power to the world economy. Nuclear energy today provides over a third of the world’s low-carbon electricity. For nuclear energy to continue to play its role in a sustainable global energy supply, both technical and institutional innovations are needed. This includes a new generation of reactors, such as advanced light water reactors, small modular reactors, including molten salt reactors and fast reactors, and even the pursuit of fusion energy.

Nuclear reactor thermal hydraulics involves the study of fluid flow, heat and mass transfer applied to nuclear technologies. It is of fundamental importance in both the design and safety operation of nuclear reactors. We are pleased to invite you to submit papers to the journal Energies for a Special Issue titled “Thermal–Hydraulic Challenges in Advanced Nuclear Reactors”. The purpose of the issue is to advance our understanding of flow and heat transfer phenomena in nuclear reactor systems to support new-generation reactor design as well as the safety of existing reactors. Experiments, system code analyses, and CFD simulations are all welcome.

The core topics include but are not limited to:

Single- and two-phase phenomena (convective heat transfer, boiling, onset of flow instability, flow regimes, single- and two-phase pressure drop);
Enhancement of boiling heat transfer;
Interphase transfer processes in two-phase flow;
Transport of radioactive trace species and aerosols in bubbles;
Condensation in two-phase flow systems with noncondensables;
Hydrodynamics of countercurrent two-phase flow;
Hydrodynamics of three-phase flow systems.

Prof. Dr. Naihua Wang
Guest Editor

Manuscript Submission Information

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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

boiling
CFD
condensation
countercurrent flow
experiment
flow instability
heat transfer
interphase transfer
mixing
nuclear reactor
safety
stratification
subchannel
two-phase flow
verification and validation

Published Papers (7 papers)

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Research

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17 pages, 6891 KiB

Open AccessArticle

Numerical Simulation of Flow and Heat Transfer Characteristics in Non-Closed Ring-Shaped Micro-Pin-Fin Arrays

and

Energies 2023, 16(8), 3481; https://doi.org/10.3390/en16083481 - 16 Apr 2023

Viewed by 1165

Abstract

In this study, flow and heat transfer characteristics in novel non-closed 3/4 ring-shaped micro-pin-fin arrays with in-line and staggered layouts were investigated numerically. The flow distribution, wake structure, vorticity field and pressure drop were examined in detail, and convective heat transfer features were explored. Results show that vortex pairs appeared earlier in the ring-shaped micro-pin-fin array compared with the traditional circular devices. Pressure drop across the microchannel varied with layout of the fins, while little difference in pressure drop was observed between ring-shaped and circular fins of the same layouts, with the maximum difference being 1.43%. The staggered ring-shaped array was found to outperform the in-line array and the circular arrays in convective heat transfer. A maximum increase of 21.34% in heat transfer coefficient was observed in the ring-shaped micro-pin-fin array in comparison with the circular micro-pin-fin array. The overall thermal-hydraulic performance of the microstructure was evaluated, and the staggered ring-shaped array with a fin height of 0.5 mm exhibited the best performance among the configurations studied. Full article

(This article belongs to the Special Issue Thermal-Hydraulic Challenges in Advanced Nuclear Reactors)

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31 pages, 7154 KiB

Open AccessFeature PaperArticle

Demonstration of Pronghorn’s Subchannel Code Modeling of Liquid-Metal Reactors and Validation in Normal Operation Conditions and Blockage Scenarios

Vasileios Kyriakopoulos

Mauricio E. Tano

and

Aydin Karahan

Energies 2023, 16(6), 2592; https://doi.org/10.3390/en16062592 - 09 Mar 2023

Cited by 1 | Viewed by 1106

Abstract

Pronghorn-SC is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE). Initially designed to simulate flows in water-cooled, square lattice, subchannel assemblies, Pronghorn-SC has been expanded to simulate liquid-metal-cooled flows in triangular lattices, hexagonal subchannel assemblies. For this purpose, the algorithm of Pronghorn-SC was adapted to solve the subchannel equations as they are applicable to a hexagonal wire-wrapped sodium-cooled fast reactor. Cheng–Todreas models for pressure drop and cross-flow models were adopted and a coolant heat conduction term was added. To solve these equations, an improved implicit algorithm was developed robust enough to deal with the numerical issues, associated with low flow and recirculation phenomena. To confirm the prediction capability of Pronghorn-SC, calculations and comparisons with available experimental data of 19- and 37-pin assemblies were performed, as well as other subchannel codes. Finally, a flow blockage modeling feature was added. This capability was validated for both water-cooled square sub-assemblies and sodium-cooled hexagonal sub-assemblies, using experimental data of partially and fully blocked cases. Full article

(This article belongs to the Special Issue Thermal-Hydraulic Challenges in Advanced Nuclear Reactors)

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16 pages, 4949 KiB

Open AccessArticle

Countercurrent Flow Limitation in a Pipeline with an Orifice

and

Energies 2023, 16(1), 222; https://doi.org/10.3390/en16010222 - 25 Dec 2022

Viewed by 951

Abstract

Countercurrent flow limitation (CCFL) refers to an important class of gravity-induced hydrodynamic processes that impose a serious restriction on the operation of gas–liquid two-phase systems. In a nuclear power plant, CCFL may occur in the liquid level measurement system where an orifice is applied in the pipeline, which may introduce error into the level measurement system. CCFL can occur in horizontal, vertical, inclined, and even much more complicated geometric patterns, and the hot-leg channel flow passage has been widely investigated; however, a pipeline with variable cross-sections, including an orifice, has not yet been investigated. An experimental investigation has been conducted in order to identify the phenomenon, pattern, and mechanism of CCFL onset in this type of geometry. Both visual and quantified experiments were carried out. A high-speed camera was applied to capture the flow pattern. Visual experiments were implemented at atmospheric pressure, while quantified pressurizer experiments were implemented at higher pressures. It was determined that if the condensate drainage is low and the liquid level is also low, with a stable stratified flow upstream of the orifice, there is no oscillation of the differential pressure. However, at higher condensate drainage levels, when the liquid level increases, a stratified wavy flow occurs. One of these waves can suddenly rise upstream of the orifice to choke it, which subsequently gives rise to differential pressure across the orifice, with periodic variation. This pattern alternately features stratified flow, stratified wavy flow, and slug flow, which indicates the occurrence of CCFL. The CCFL occurring under these experimental conditions can be expressed as a Wallis type correlation, where the coefficients m and C are 0.682 and 0.601, respectively. Full article

(This article belongs to the Special Issue Thermal-Hydraulic Challenges in Advanced Nuclear Reactors)

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18 pages, 3245 KiB

Open AccessArticle

The Effect of Nodalization Schemes on the Stability Characteristics of a Three Heated Channels under Supercritical Flow Condition

Munendra Pal Singh

Abdallah Sofiane Berrouk

and

Muhammad Saeed

Energies 2022, 15(23), 9046; https://doi.org/10.3390/en15239046 - 29 Nov 2022

Viewed by 1028

Abstract

The present analysis is aimed at conducting node sensitivity analysis on the thermal–hydraulic performance of supercritical fluid in a three parallel channel configuration system. The heated channel was divided into different nodes and is examined under wide-ranging operating conditions. Firstly, the heated channel was divided into two nodes, like the two-phase flow system. In the second case, based on the correlation between the fluid properties, the heated channel was divided into three regions: heavy, mixture, and supercritical fluids. Finally, the channel was divided into N-nodes. Post the nodalization analysis, a non-linear analysis of three parallel channels was carried out under varied heat flux conditions. The analytical approximation functions were developed to capture the fluid flow dynamics. These functions were used to capture each node’s density, enthalpy, and velocity profiles under a wide range of operating conditions. The different flow instability (density wave oscillations and static) characteristics were observed at low pseudo- and relatively high subcooling numbers. In the density wave oscillations regime, out-of-phase oscillations and limit cycles are observed. A co-dimension parametric analysis with numerical simulations was carried out to confirm the obtained non-linear characteristics. Such analysis for parallel channel systems under supercritical working fluid flow conditions is missing in the literature which is limited to only linear stability analysis. This analysis can help to improve heat and mass transfer for designing efficient heated channel systems. Full article

(This article belongs to the Special Issue Thermal-Hydraulic Challenges in Advanced Nuclear Reactors)

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15 pages, 4576 KiB

Open AccessArticle

The Investigation of the Bubble Behaviors on the Vertical Heat Exchange Tube

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Energies 2022, 15(19), 7097; https://doi.org/10.3390/en15197097 - 27 Sep 2022

Viewed by 849

Abstract

In the boiling process, the growth, separation, and movement of bubbles are expeditious. The visualization experiment of nucleate boiling was carried out with the help of high-speed photography. The evolution of the entire bubble life cycle is clearly observed at the nucleation site without interference from the leading and neighboring bubbles. Bubble behavior at the local heating surface has strong randomness due to the influence of the wall micro-structure, convection intensity, heating surface geometry configuration, heat flux density, and so on, but bubble behavior also has a certain regularity. In this paper, the behavior characteristics of bubbles were analyzed, with a particular focus on the evolution of bubbles. Under lower load (ΔT_sat = 8~9 °C) in study conditions, nucleation sites have a long enough time interval. In addition, the bubble separation and rising velocity obviously increase due to the change of pool boiling flow characteristics in the restricted space. The setting of confined space increases the bubble escape velocity and the rising velocity, and decreases the diameter of bubbles escaping from the wall. The results will provide some help for the understanding of bubble behavior mechanisms and numerical research. Full article

(This article belongs to the Special Issue Thermal-Hydraulic Challenges in Advanced Nuclear Reactors)

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Review

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20 pages, 5623 KiB

Open AccessReview

Research Progress on Thermal Hydraulic Characteristics of Spent Fuel Pools: A Review

and

Energies 2023, 16(10), 3990; https://doi.org/10.3390/en16103990 - 09 May 2023

Viewed by 1063

Abstract

Nuclear power plants (NPPs) produce large amounts of spent fuel while generating electricity. After the spent fuel is taken out of the reactor core, it still has a high decay heat and needs to be cooled for years or even decades before it can be reprocessed or buried deeply. Due to the long storage period of spent fuel, storage safety evaluation is a concern. In this regard, cooling systems are critical for the safe storage of spent fuel. Here, the research progress of cooling methods for spent fuel pools (SFPs) is reviewed, and the structural characteristics, application limitations and heat transfer performance of active and passive cooling technologies under accident conditions are discussed in detail. Moreover, future developments of SFPs are discussed, and the results of this review confirm that there is a great deal of research scope to improve the cooling performance and safety of spent fuel. This paper aims to provide a reference guide for engineers and will be highly beneficial to researchers engaged in spent fuel storage. Full article

(This article belongs to the Special Issue Thermal-Hydraulic Challenges in Advanced Nuclear Reactors)

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43 pages, 5873 KiB

Open AccessReview

Experimental Investigations and Numerical Studies of Two-Phase Countercurrent Flow Limitation in a Pressurized Water Reactor: A Review

and

Energies 2023, 16(3), 1487; https://doi.org/10.3390/en16031487 - 02 Feb 2023

Cited by 2 | Viewed by 1305

Abstract

Gas–liquid two-phase countercurrent flow limitation (CCFL) phenomena widely exist in nuclear power plants. In particular, the gas–liquid countercurrent flow limitation phenomena in a pressurized water reactor (PWR) during a loss-of-coolant accident (LOCA) or a small-break loss-of-coolant accident (SBLOCA) play an important role in nuclear reactor safety research. Over several decades, a series of experimental investigations and numerical studies have been carried out to study the CCFL phenomena in a PWR. For the experimental investigations, numerous experiments have been conducted, and different CCFL mechanisms and CCFL characteristics have been obtained in various test facilities simulating different scenarios in a PWR. The CCFL phenomena are affected by many factors, such as geometrical characteristics, liquid flow rates, and fluid properties. For the numerical studies, more and more numerical models were presented and applied to the calculations of two-phase countercurrent flow over the past several decades. It is considered that the computational fluid dynamics (CFD) tools can simulate most of the two-phase flow configurations encountered in nuclear power plants. In this paper, the experimental investigations and the numerical studies on two-phase countercurrent flow limitation in a PWR are comprehensively reviewed. This review provides a further understanding of CCFL in a PWR and gives directions regarding future studies. It is found that relatively fewer investigations using steam–water under high system pressures are performed due to the limitation of the test facilities and test conditions. There are a number of numerical studies on countercurrent two-phase flow in a PWR hot leg geometry, but the simulations in other flow channels were relatively rare. In addition, almost all of the numerical simulations do not include heat and mass transfer. Thus, it is necessary to investigate the effects of heat and mass transfer experimentally and numerically. Furthermore, it is of significance to perform numerical simulations for countercurrent two-phase flow with a fine computational grid and suitable models to predict the formation of small waves and the details in two-phase flow. Full article

(This article belongs to the Special Issue Thermal-Hydraulic Challenges in Advanced Nuclear Reactors)

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