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Optimal Design and Analysis of Advanced Nuclear Reactors
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Optimal Design and Analysis of Advanced Nuclear Reactors

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

Deadline for manuscript submissions: 10 March 2025 | Viewed by 6642

Special Issue Editors

Dr. Jing Zhang

E-Mail Website
Guest Editor
Department of Nuclear Science and Technology, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an 710049, China
Interests: nuclear reactor design; safety and simulation of nuclear power system; nuclear reactor thermal hydraulic; artificial intelligence
Dr. Yuan Yuan

E-Mail Website
Guest Editor
College of Physical Science and Technology, Sichuan University, Chengdu 610065, China
Interests: reactor thermal hydraulic; reactor numerical calculation; artificial intelligence

Special Issue Information

Dear Colleagues,

Nuclear energy is an efficient and clean type of energy. The nuclear reactor-based energy supply is characterized by high energy density, low carbon emissions, long sustainable operation time and wide use. Reactor technology continues to evolve, with a large number of passive generation III+ and generation IV reactor designs emerging. All designs focus on the reactor's inherent safety improvement, while allowing for a more efficient and flexible energy supply. With the development of experimental measurement and computer simulation technology, more accurate analytical methods provide support for the design and optimization of advanced reactors; thus, the economy and safety of reactors would be enhanced.

This Special Issue aims to present and disseminate the most recent advances related to the theory, design, modeling and optimization of all types of advanced nuclear reactors. Topics of interest for publication include, but are not limited to:

  • Thermal-hydraulic characteristics of advanced reactors;
  • Multi-physics coupling in the reactor core;
  • Nuclear reactor systems design;
  • Safety analysis of advanced reactors;
  • Explicable machine learning in nuclear energy;
  • Advanced optimization algorithms.

Dr. Jing Zhang
Dr. Yuan Yuan
Guest Editors

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

  • thermal hydraulics
  • safety analysis
  • multi-physics coupling
  • explicable machine learning in nuclear energy
  • severe accident

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Published Papers (5 papers)

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Research

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38 pages, 14509 KiB  
Article
Studying Dynamical and Hydraulic Characteristics of the Hydraulically Suspended Passive Shutdown Subassembly (HS-PSS) and Validating with a Prototypic Test Sample
by Bo Kuang, Xin Wang, Jieming Hou and Wenjun Hu
Energies 2024, 17(20), 5038; https://doi.org/10.3390/en17205038 - 10 Oct 2024
Viewed by 500
Abstract
The pool-type demonstrative China Fast Reactor 600 (CFR-600) adopted a series of improved safety designs, among which the hydraulic suspended passive shutdown subassembly (HS-PSS) is employed for inherent safety enhancement, especially suitable against unprotected loss of flow accident (ULOF). In this article, functional [...] Read more.
The pool-type demonstrative China Fast Reactor 600 (CFR-600) adopted a series of improved safety designs, among which the hydraulic suspended passive shutdown subassembly (HS-PSS) is employed for inherent safety enhancement, especially suitable against unprotected loss of flow accident (ULOF). In this article, functional requirements for HS-PSS design in hydro-dynamic aspects are proposed, with the corresponding performance indicators discussed. To address these functional requirements, qualitative analysis on the equilibrium solution properties of the nonlinear dynamical model equation of the hydraulic moving body (HMB) in the HS-PSS are conducted, which leads to the determination of an applicable design parameter domain of the HMB for its practical design from the broad range of structural and parametric design options available for HS-PSS. Furtherly, hydraulically characterizing and modeling the constituent paths and consequent fluid network, hydraulic characteristics of the HS-PSS, as well as the coupled hydro-dynamic motion behaviors of the HMB for suspension and dropping states, were simulated and test-validated. Considering the HS-PSS hydro-dynamic behaviors, key indicators such as critical flowrate, drop time, hydraulic self-tightening performance, as well as the hydraulic characteristics curve are for fulfilling the functional requirements. Meanwhile, through sensitivity study of some structural parameters‘ impact on hydraulic characteristics, some most sensible structural parameters for adjusting and optimizing detailed design are observed. The work is quite significant in supporting the conceptual design of the HS-PSS as well as its engineering improvement. Full article
(This article belongs to the Special Issue Optimal Design and Analysis of Advanced Nuclear Reactors)
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25 pages, 8124 KiB  
Article
Study of Condensation during Direct Contact between Steam and Water in Pressure-Relief Tank
by Shasha Yin, Yingjie Wang, Yuan Yuan and Bei Li
Energies 2024, 17(11), 2772; https://doi.org/10.3390/en17112772 - 5 Jun 2024
Viewed by 1136
Abstract
Direct contact condensation (DCC) is a phenomenon observed when steam interacts with subcooled water, exhibiting higher heat and mass transfer rates compared to wall condensation. It has garnered significant interest across industries such as nuclear, chemical, and power due to its advantageous characteristics. [...] Read more.
Direct contact condensation (DCC) is a phenomenon observed when steam interacts with subcooled water, exhibiting higher heat and mass transfer rates compared to wall condensation. It has garnered significant interest across industries such as nuclear, chemical, and power due to its advantageous characteristics. In the context of pressure-relief tanks, understanding and optimizing the DCC process are critical for safety and efficiency. The efficiency of pressure-relief tanks depends on the amount of steam condensed per unit of time, which directly affects their operational parameters and design. This study focuses on investigating the direct gas–liquid contact condensation process in pressure-relief tanks using computational fluid dynamics (CFD). Through experimental validation and a sensitivity analysis, the study provides insights into the influence of inlet steam parameters and basin temperature on the steam plume characteristics. Furthermore, steady-state and transient calculation models are developed to simulate the behaviour of the pressure-relief tank, providing valuable data for safety analysis and design optimization. There is a relatively high-pressure area in the upper part of the bubble hole of the pressure-relief tube, and the value increases as it is closer to the holes. The steam velocity in the bubbling hole near the 90° elbow position is higher. This study contributes to the understanding of steam condensation dynamics in pressure-relief tanks. When the steam emission and pressure are fixed, the equilibrium temperature increases linearly as the initial temperature increases (where a = 1, b = 20 in y = a x+ b correlation), the equilibrium pressure increases nearly exponentially, and the equilibrium gas volume decreases. When the steam emission and initial temperature are fixed, the equilibrium temperature does not change as the steam discharge pressure increases. The correlations between the predicted equilibrium parameters and the inlet steam parameters and tank temperature provide valuable insights for optimizing a pressure-relief tank design and improving the operational safety in diverse industrial contexts. Full article
(This article belongs to the Special Issue Optimal Design and Analysis of Advanced Nuclear Reactors)
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14 pages, 2948 KiB  
Article
Coupling Design and Validation Analysis of an Integrated Framework of Uncertainty Quantification
by Bo Pang, Yuhang Su, Jie Wang, Chengcheng Deng, Qingyu Huang, Shuang Zhang, Bin Wu and Yuanfeng Lin
Energies 2023, 16(11), 4435; https://doi.org/10.3390/en16114435 - 31 May 2023
Viewed by 1090
Abstract
The uncertainty quantification is an indispensable part for the validation of the nuclear safety best-estimate codes. However, the uncertainty quantification usually requires the combination of statistical analysis software and nuclear reactor professional codes, and it consumes huge computing resources. In this paper, a [...] Read more.
The uncertainty quantification is an indispensable part for the validation of the nuclear safety best-estimate codes. However, the uncertainty quantification usually requires the combination of statistical analysis software and nuclear reactor professional codes, and it consumes huge computing resources. In this paper, a design method of coupling interface between DAKOTA Version 6.16 statistical software and nuclear reactor professional simulation codes is proposed, and the integrated computing workflow including interface pre-processing, code batching operations, and interface post-processing can be realized. On this basis, an integrated framework of uncertainty quantification is developed, which is characterized by visualization, convenience, and efficient computing. Meanwhile, a typical example of small-break LOCA analysis of the LOBI test facility was used to validate the reliability of the developed integrated framework of uncertainty quantification. This research work can provide valuable guidance for developing an autonomous uncertainty analysis platform in China. Full article
(This article belongs to the Special Issue Optimal Design and Analysis of Advanced Nuclear Reactors)
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18 pages, 10356 KiB  
Article
Three-Dimensional Surrogate Model Based on Back-Propagation Neural Network for Key Neutronics Parameters Prediction in Molten Salt Reactor
by Xinyan Bei, Yuqing Dai, Kaicheng Yu and Maosong Cheng
Energies 2023, 16(10), 4044; https://doi.org/10.3390/en16104044 - 12 May 2023
Cited by 2 | Viewed by 1439
Abstract
The simulation and analysis of neutronics parameters in Molten Salt Reactors (MSRs) is fundamental for the design of the reactor core. However, high-fidelity neutron transport calculations of the MSR are time-consuming and require significant computational resources. Artificial neural networks (ANNs) have been used [...] Read more.
The simulation and analysis of neutronics parameters in Molten Salt Reactors (MSRs) is fundamental for the design of the reactor core. However, high-fidelity neutron transport calculations of the MSR are time-consuming and require significant computational resources. Artificial neural networks (ANNs) have been used in various industries, and in recent years are increasingly introduced in the nuclear industry. Back-Propagation neural network (BPNN) is one type of ANN. A surrogate model based on BP neural network is developed to quickly predict two key neutronics parameters in reactor core: the effective multiplication factor (keff) and the three-dimensional channel-by-channel neutron flux distribution. The dataset samples are generated by modeling and simulating different operation states of the Molten Salt Reactor Experiment (MSRE) using the Monte Carlo code. Hyper-parameters optimization is performed to obtain the optimal surrogate model. The numerical results on the test dataset show good agreement between the surrogate model and the Monte Carlo code. Additionally, the surrogate model significantly reduces computation time compared to the Monte Carlo code and greatly enhances efficiency. The feasibility and advantages of the proposed surrogate model is demonstrated, which has important significance for real-time prediction and design optimization of the reactor core. Full article
(This article belongs to the Special Issue Optimal Design and Analysis of Advanced Nuclear Reactors)
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Review

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23 pages, 2961 KiB  
Review
Research Advances in the Application of the Supercritical CO2 Brayton Cycle to Reactor Systems: A Review
by Yuhui Xiao, Yuan Zhou, Yuan Yuan, Yanping Huang and Gengyuan Tian
Energies 2023, 16(21), 7367; https://doi.org/10.3390/en16217367 - 31 Oct 2023
Viewed by 1675
Abstract
Amid the global emphasis on efficient power conversion systems under the “dual carbon” policy framework, the supercritical CO2 (SCO2) Brayton cycle is a noteworthy subject, owing to its pronounced efficiency, compact design, economic viability, and remarkable potential to increase the [...] Read more.
Amid the global emphasis on efficient power conversion systems under the “dual carbon” policy framework, the supercritical CO2 (SCO2) Brayton cycle is a noteworthy subject, owing to its pronounced efficiency, compact design, economic viability, and remarkable potential to increase the thermal cycle efficiency of nuclear reactors. However, its application across various nuclear reactor loops presents divergent challenges, complicating system design and analytical processes. This paper offers a thorough insight into the latest research on the SCO2 Brayton cycle, particularly emphasising its integration within directly and indirectly cooled nuclear reactors. The evolution of the Brayton cycle in nuclear reactor systems has been meticulously explored, focusing on its structural dynamics, key components, and inherent pros and cons associated with distinct reactor loops. Based on the theoretical frameworks and empirical findings related to turbomachinery and heat exchangers within the cycle, we chart a course for future enquiries into its critical components, underscoring the indispensable role of experimental investigations. This paper conclusively assesses the feasibility of deploying the SCO2 Brayton cycle in direct and indirect cooling contexts, offering a forward-looking perspective on its developmental trajectory. The SCO2 Brayton cycle may become a focal point for research, potentially creating avenues for nuclear energy endeavours. Full article
(This article belongs to the Special Issue Optimal Design and Analysis of Advanced Nuclear Reactors)
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