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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: 18 June 2024 | Viewed by 6787

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Special Issue Editor

Prof. Dr. Naihua Wang

E-Mail Website
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 (5 papers)

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Research

13 pages, 4174 KiB

Open AccessArticle

Experimental Investigation on Pressure Drops of Purge Gas Helium in Packed Pebble Beds for Nuclear Fusion Blanket

by Hao Cheng, Baoping Gong, Bing Zhou, Juemin Yan, Long Wang, Long Zhang, Yongjin Feng and Xiaoyu Wang

Energies 2024, 17(6), 1309; https://doi.org/10.3390/en17061309 - 8 Mar 2024

Viewed by 480

Abstract

The flow characteristics of purge gas helium in the pebble bed of the tritium breeding blanket are important in analyzing the tritium purging process and optimizing the design of the solid breeder blanket. Therefore, the flow characteristics of helium gas in randomly packed pebble beds are investigated experimentally with a focus on the analysis of the pressure loss distribution. The results show that gas velocity, bed dimension, and pebble diameters have an obvious influence on the helium flow characteristics in pebble beds. With the increase in the inlet helium gas velocity, the pressure drop gradient of helium in the pebble bed gradually increased. With increases in the pebble bed dimension, the pressure drop gradient of helium in the pebble bed gradually increased. With the increase in the pebble diameter, the pressure drop gradient gradually decreased. In addition, the effect of temperature on the pressure drop of helium in the pebble bed was also preliminarily investigated. The pressure drop gradient of the helium through the pebble bed obviously increased with the increase in the helium and the bed temperature. The experimentally obtained pressure loss characteristics can be used for the validation of the simulation of a blanket pebble bed and as input parameters in the thermo-hydraulic analysis of solid-tritium breeder blankets. Full article

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

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Graphical abstract

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

by Vasileios Kyriakopoulos, Mauricio E. Tano and Aydin Karahan

Energies 2023, 16(6), 2592; https://doi.org/10.3390/en16062592 - 9 Mar 2023

Cited by 3 | Viewed by 1568

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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Graphical abstract

16 pages, 4949 KiB

Open AccessArticle

Countercurrent Flow Limitation in a Pipeline with an Orifice

by Danni Zhao, Chende Xu, Zhengguang Wang, Xixi Zhu, Yaru Li, Xiangyu Chi and Naihua Wang

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

Cited by 2 | Viewed by 1326

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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Graphical abstract

18 pages, 3245 KiB

Open AccessArticle

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

by 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 1367

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

by Yongsheng Tian, Pengfei Xu, Linhua Zhang and Luopeng Yang

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

Viewed by 1169

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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Planned Papers

The below list represents only planned manuscripts. Some of these manuscripts have not been received by the Editorial Office yet. Papers submitted to MDPI journals are subject to peer-review.

Title: Numerical simulation of heat transfer performance of an axially grooved heat pipe
Authors: Kaimin Yang
Affiliation: Shandong Jianzhu University

Title: Adaptive coordinate control of NSSS of FHR
Authors: Yiran LIU
Affiliation: China General Nuclear Power Group

Title: Modeling method of once-though steam generator thermos hydraulic transfer function of compact spiral pipe based on experiment
Authors: Tianming Chen
Affiliation: China General Nuclear Power Group

Title: Analysis of coolant flow distribution at the core inlet for a small Pressurized Water Reactor in ocean motion
Authors: Jinyuan Zeng
Affiliation: China General Nuclear Power Group

Title: CFD analysis of the effect of fuel rod growth of a PWR fuel assembly inlet region
Authors: Kejia LI
Affiliation: SUN YAT-SEN UNIVERSITY

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