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Optimal Design and Analysis of 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 (10 March 2025) | Viewed by 11151

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Special Issue Editors

Dr. Jing Zhang

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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

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Keywords

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

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Published Papers (9 papers)

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Research

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21 pages, 18800 KiB

Open AccessArticle

Research on Thermo-Mechanical Response of Solid-State Core Matrix in a Heat Pipe Cooled Reactor

by Xintong Peng, Cong Liu, Yangbin Deng, Jingyu Nie, Yingwei Wu and Guanghui Su

Energies 2025, 18(6), 1423; https://doi.org/10.3390/en18061423 - 13 Mar 2025

Viewed by 393

Abstract

Due to its advantages of simple structure and high inherent safety, the heat pipe cooled reactor (HPR) could be widely applied in deep-sea navigation, deep-space exploration and land-based power supply as a promising advanced special nuclear power equipment option. In HPRs, the space between the components (fuel rods and heat pipes) is filled with solid matrix material, forming a continuous solid reactor core. Thermo-mechanical response of the solid core is a special issue for HPRs and has great impacts on reactor safety. Considering the irradiation and burnup effects, the thermal and mechanical modeling of an HPR was conducted with ABAQUS-2021 in this study. The thermo-mechanical response under long-term normal operation, accident transients and single heat pipe failed conditions was simulated and analyzed. The whole core presents relatively good isothermality due to the high thermal conductivity of the solid matrix. As for the mechanical performance, the maximum stress was about 300 MPa, and the maximum displacement of the matrix could be as high as 3.7 mm. It could lead to significant variation of the reactor physical parameters, which warrants further attention in reactor design and safety analysis. Reactivity insertion accidents or single heat pipe failure has obvious influence on the thermo-mechanical performance of the local matrix, but they did not cause any failure risks, because the HPR design eliminates the dramatic power flash-up and the solid-state core avoids the heat transfer crisis caused by the coolant phase transition. A quantitative evaluation of thermo-mechanical performance was completed by this research, which is of great value for reactor design and safety evaluation of HPRs. Full article

(This article belongs to the Special Issue Optimal Design and Analysis of Advanced Nuclear Reactors)

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15 pages, 2850 KiB

Open AccessArticle

Study on Reactor Power Control Strategies Based on Mode-C Operation and Control Mode

by Ying Zhang, Zhi Chen, Qing Chu and Jixiang Zhou

Energies 2025, 18(5), 1140; https://doi.org/10.3390/en18051140 - 26 Feb 2025

Viewed by 351

Abstract

At present, the operation control modes of pressurized water reactor (PWR) nuclear power plants in service mainly include Mode-A, Mode-G, and MSHIM. Mode-A is mainly applicable to base load operation and cannot realize load tracking. In the process of Mode-G load tracking, it is necessary to adjust boron, and it cannot realize load tracking without boron regulation. Although MSHIM implements unregulated boron load tracking, a large number of control rods are inserted into the core during base load operation, which reduces the safety margin and causes certain economic losses. In recent years, China National Nuclear Corporation Limited proposed the Mode-C operation control mode, which attempts to concentrate the advantages of the above operation mode and avoid its disadvantages. When Mode-C is adopted, only one set of control rods is inserted into the reactor core to complete the nuclear power plant control task for the base load and other operations that do not require frequent reactor power regulation. For load tracking and other operations requiring frequent reactor power regulation, control rods are used instead of adjusting soluble boron to control core reactivity. Reactivity compensation and power distribution control in the load-tracking process are completed through control rods. When Mode-C mode is adopted, the reactivity control method under base load and load tracking conditions is different from other mature operating modes. It is impossible to directly adopt the ready-made reactor power control system scheme, which brings challenges to the practical engineering application of Mode-C. To solve the above problems, based on the idea of single-variable automatic control and bivariable automatic control in Mode-C under different load demand conditions, this paper carries out research on the strategy of the reactor power control system and puts forward two specific control schemes. Through the control simulation program based on the one-dimensional core model, the simulation model of the control object and control system is established, and the closed-loop simulation verification of the control strategy is completed. The simulation results show that the designed reactor power control system can realize automatic control of the full power operating range and non-adjustable boron load tracking, reduce the operator’s burden, and meet the expected operation effect of the Mode-C operating mode. Full article

(This article belongs to the Special Issue Optimal Design and Analysis of Advanced Nuclear Reactors)

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25 pages, 7066 KiB

Open AccessArticle

Dynamic Effect of the Delayed Neutron Precursor Distribution on System Safety Analysis in Liquid-Fueled Molten Salt Reactor

by Shichao Chen, Rui Li, Xiandi Zuo, Maosong Cheng and Zhimin Dai

Energies 2025, 18(3), 670; https://doi.org/10.3390/en18030670 - 31 Jan 2025

Viewed by 752

Abstract

The liquid-fueled molten salt reactor (MSR) is one of the candidate reactors for the Generation IV advanced nuclear power systems, which utilizes flowing liquid molten salt as both fuel and coolant. In transients of liquid-fueled MSRs, the distribution change in the delayed neutron precursors (DNPs) in the primary loop has an important impact on system safety analysis. In order to analyze and evaluate this effect, the RELAP5-TMSR code with a 1-D DNP transport model was used to model the Molten Salt Breeder Reactor (MSBR), and several representative transient scenarios, including the loss of primary flow, increase in primary flow, loss of secondary flow, reactivity perturbation, and load change, were simulated and analyzed. The results show that the DNP distribution changes obviously during primary flow transients, especially during the loss of primary flow. Besides, the power response trends at different power levels during the loss of primary flow are different. The analysis results reveal the steady-state and dynamic characteristics of the DNP distribution, indicating that the DNP distribution, temperature feedback, and reactor power are strongly coupled, which has significant implications for the design and safety analysis of liquid-fueled MSRs. Full article

(This article belongs to the Special Issue Optimal Design and Analysis of Advanced Nuclear Reactors)

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22 pages, 12290 KiB

Open AccessArticle

Enhancing Thermal-Hydraulic Modelling in Dual Fluid Reactor Demonstrator: The Impact of Variable Turbulent Prandtl Number

by Hisham Elgendy, Sławomir Kubacki and Konrad Czerski

Energies 2025, 18(2), 396; https://doi.org/10.3390/en18020396 - 17 Jan 2025

Viewed by 696

Abstract

In response to the growing demand for advanced nuclear reactor technologies, this study addresses significant gaps in thermal-hydraulic modelling for dual fluid reactors (DFRs) by integrating Kays correlation to implement a variable turbulent Prandtl number in the Reynolds-averaged Navier–Stokes (RANS) simulations. Traditional approaches employing a constant value of the turbulent Prandtl number have proven inadequate, leading to inaccurate heat transfer predictions for low Prandtl number liquids. The study carefully selects the appropriate formula for the turbulent Prandtl number in the DFR context, enhancing the accuracy of thermal-hydraulic modelling. The simulations consider Reynolds numbers between 15,000 and 250,000, calculated based on the hydraulic diameters at different diameter pipes of the fuel and coolant loops. The molecular Prandtl number is equal to 0.025. Key findings reveal that uneven flow distributions within the fuel pipes result in variable temperature distribution throughout the reactor core, confirming earlier observations while highlighting significant differences in parameter values. These insights underscore the importance of model selection in CFD analysis for DFRs, revealing potential hotspots and high turbulence areas that necessitate further investigation into vibration and structural safety. The results provide a framework for improving reactor design and operational strategies, ensuring enhanced safety and efficiency in next-generation nuclear systems. Future work will apply this modelling approach to more complex geometries and flow scenarios to optimise thermal-hydraulic performance. Full article

(This article belongs to the Special Issue Optimal Design and Analysis of Advanced Nuclear Reactors)

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38 pages, 14509 KiB

Open AccessArticle

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

Cited by 1 | Viewed by 912

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 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

Open AccessArticle

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 1643

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. 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

Open AccessArticle

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 1246

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 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

Open AccessArticle

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 3 | Viewed by 1790

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 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 (

k_{eff}

) 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

Open AccessReview

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

Cited by 5 | Viewed by 2308

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 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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